ML20115K089

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Second Quarter 1992,April-June

ML20115K089
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Issue date:	09/30/1992
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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NUREG-0304, NUREG-0304-V17-N02, NUREG-304, NUREG-304-V17-N2, NUDOCS 9210290091
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1 NUREG-0304 Vol.17, No. 2

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Regu:a:ory anc Technica: Reports (Abs:rac: Inc ex JournaD Compilation for Second Quarter 1992 April - June U.S. Nucler Regulatory Commission Omce of Administration

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Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C, 20013 7082 A year's subscription consists of 4 issues for this publication, Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 p

NUREG-0304 Vol.17, No. 2

.==

. Regulatory and Technical Reaorts (Abstract Index Journal)

Compilation for Second Quarter 1992 April - June Date Published: September 1992 Regulatory Publications llranch Division of Freedom ofInformation and Publications Services OITice of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 y...m,,

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CONTENTS Preface.

v index Tab Main Citations and Abstracts

.1

• Staff Reports a Conference Proceedings

• Contractor Report?

• International Agreement Reoorts Secondary Report Number Index.

2 Personal Author index.

3 Subject index....

.4 NRC Originating Organization index (Staff Reports)..

5 NRC Originating Organization index (International Agreements)..

6 NRC Contract Sponsor index (Contractor Reports),

.7 Contractor index.,...

8 International Organization index.,

....9 Licensed Facility index...

.. 10 1

PREFACE This compilation consists of bibliographic data and abstracts for the formal regulatory and technical.

reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. it is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments wil' be ag.

. preciated. Please send them to:

Technical Publications Section Regulatory P"blications Branch Division of Freedom of Information and Publications Services P-223 U.S. Nuclear Regulatory Commission Washin0 ton. D.C. 20555 The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/lA-XXXX. These precede the following indexes:

Secondary Report Number Index Personal Author index-Subject index NRC Originating Organization Index (Staff Reports)

NRC Originating Organization Index (International Agreements)

NRC Contract Sponsor index (Contractor Reports)

Contractor Index Internatic,nal Organization index Licensed Facility Index A detailed explanation nf the entries precedes each iridex.

The bibliographic elements of the main citations are the following:

Staff Report NUREG-0808: MARK ll CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA.

ANDERSON, C.J. Division of Safety Technology. August 1981. 00 pp.' 8109140048 09570:200.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the micro'iche address (for internal NRC use).

(-

Conference Report i.

l NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT AND l

RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National L

Laboratory. May 1981.141 pp. 8105280299. ANL-81-3, 08832:070, l

Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled the proceedings, (5) date report was published. (6) number of pages in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC internal use).

Contractor Report l

NUREG/CR-1E"d: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.

l Sandia Laboratories. May 1981.100 pp. 8107010449. SAND 80-0029. 08912:242.

I Where the entries are (1) report number, (2) report title, (3) report authors, (4) organizational unit of l

authors or publisher, (5) date report was published. (6) number of pages in the report, (7) the NRC i

Document Control System accession number, (8) the report number of the originating organization (if I

given), and (9) the microfiche address (for NRC internal use).

v t

I

I ternational Agreement ReVrt f

NUREG/lA 0001: ASSESSMENT OF TRAC-PD2 JSING SUPER CANNON AND HDR EXPERIMENTAL DATA NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424, 37659:138.

Where the entries are (1) report number. (2' report title, (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and (9) the mictrMche address (for NRC internal use).

The following abbreviations are used to identify the document status of a reptrt:

ADD

- addendum APP

- appendix DRFT - draf t ERR

- errata N

number i

R - revision S

supplement V - volume Availabi:ity of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Govemment Printing Office (GPO) or from the National Technical Information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the followin0 address:

Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, DC X;013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275 2171. Non-U.S. customers must make payment in advance either by International Postal Money Order, payable to the Superintendent of Documents, or by draft on a United States or Canadian bank, payable to the Superintendent of Documents.

NRC F.eport Codes The NUREG designation, NUREG-XXXX, indicates that the document is.ormal NRC staff-generated report. Contractor-prepared formal NRC reports carry the report code h.uREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG XXXX, as well as various other numbers that could not be correlated veith NRC sponsorship of the work being reported.

In addition to the NUREG and NUREG/CR codes, NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/lA is used for international egreement reports.

All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Division of Pubications Services.

vi

Main Citations ard Abstracts The report listings in this compilation are arranged by report number, where NUREG XXXX is

n NRC staff-originated report, NUREG/CP-XLXX is an NRC-sponsored conference report, NUREG/CR XXXX is an NRC contractor-prepared report, and NUREG/lA-XXXX is an inter-national agreement report, The bibliographic information (see Preface for deta; is followed by a brief abstract of this report.

NUREG-0040 V16 N01: LICENSEE CONTRACTOR AND NUREG-0540 Vid NO2: TITLE LIST OF DOCUMENTS MADE VENDOR INSPECTION STATUS REPORT, Quarterty PUBLICLY AVAILABLE. February 1 29,1992.

• Divison of Free-Report.Jenuary - Mc ch 1992 (White Book)

• Division of Reactoi dom of Information & Publicatens Services (Post 890205) April inspection & Safeguards (Post 870411). May 1992. 400pp.

1992. 361pp. 9205140220. 61669:140, 9206.20324. 61998.222.

This document is a monthly publicaton containing desenp-Th a punodical covers the results of inspettons performed by tions of information received and geno ated by the U.S. Nuclear the NRC's Vendor inspection Branch that have been distnbuted Regulate y Commisstor. fNRC). This information includos* (1) te the inspected organizatens dunng the pmod from January docketed material associated with civihan nuclear power plants and other uses of rac'Oactive materials, and (2) nondocketed through March 1992.

matonal received and generated by NRC pertinent to tts role as NUREG-0000 V14 N04. RFPORT TO CONGRESS ON ABNOR, a regulatory 6gency The following indexos are included: Pet-MAL OCCURRENCES. October. December G91.

• Office for sonal Author, Corporate Sou;ce, Report Number, and Cross Reference of Enclosures to Pnncipal Documents.

Ana'ysis & Evaluation of Opcratonal Data, Director. March 1992. 31pp. 9204300058 61502112-NUREG 0540 V14 NO3: TITLE LIST OF DOCUMENTS MACC Section 208 of the Energy Reorganizaten Act of 1974 identi-PUBLICLY AVAILABLEMarch 1 31, 1992.

• Division of Free-fics an abnormal occurrenco as an unscheduled incident or dom of mformahon & Pubhcations Services (Pont 890205). May event that the Nuclear Regulatory Commission determines to be 1992. 444pp 9206030189. 61873:185.

significant from the standpoint of public health and safety and See NUREG 0540 V14,N02 abstract.

requires a quarterly report of such events to be made to Con-

'wess. This report covers the penod October through December NUREG 0540 V14 N04: TITLE LIST OF DOCUMENTS MADE 191. Five abnormal occurrences at NRC-heensed iacihties are PUBLICLY AVAILABLE.Apnl 1 30, 1992.

• Division of Freedom of ormation & Publications Cirvices (Post t :0205). June

.aussed in this rerort. None of these occurrences involved a 1992. 331pp. 9207020031. 62189.066.

nuclear power plant. Four involved medical therapy misadminis, See NUREG 0540 V14,N02 abstract.

trations and one involved a medical diagnostic misadministra-tion. The NRC's Agreement States reported three abnormal oc-e,UREG-0544 R03: NRC COLLECTION OF ABBREVIATIONS.

• currences. Two involved exposures of nnn-radiation workers Division of Freedon of Information & Pubhcations Services and one involved a -medica! therapy misadministration. The (Post 800205). March 1592,220pp. 9204210076. 61401001, report also contains inforrraton tnat updates some previously The U S. Nuclear Regulatory Comm6ssen (NRC) staff collect-reported abnormrt occurrences.

ed this hst of abbreviations from NRC documents and nuclear industry documents, both foreign and comu ' Readers can l

NUREG-0304 V16 N04: REGULATORY AND TECHNICAL RE-use the collection, which is not all inclusive, to iuentify the terms PORTS (ABSTRACT INDEX JOURNAL). Annual Compdation from which the abbreviations are formed. The Editonal Secton l

For 1991.

• Dnnvon of Freedom of informahon & Pubhcationc of the Division of Freedom of Information and Pubhcatens Serv-('

Servicos (Post 890205). March 1992. 118pp. 9204210071.

ices compiled this collection. In the introduction, the editorial 61391:238.

staff offers suggestions for using abbreviations but does not This journal includes au format reports in the NUREG senes recommend the use of one abbreviation over another.

prepared by the NRC staff and contractors; proceedings of con ~

NUREG-0713 Vt1: OCCUPATIONAL RADIATION EXPOSURE AT ferences and workshops: as well as internatv ai agreement re" COMMERCIAL NUCLEAR POWER REACTORS AND OTHER ports. The entnes in this compilation are indexed for access by FACILITIES.1989 Twonty Second Annual Repors title and abstract, secondary report number. personal autW RADDATZ,C.T. Division of Regulatory Applicatons (Post subject, NRC organtzation for staff and international agree-870413). HAGEMEYER,0. Science Apphcations Internahonal ments, contractor. international organization, and kcensed f acifi-Corp (formerly Science Apphcat ons, Inc ). Apnl 1992. 296pp.

ty.

9204230158. 61437:127.

This report summan7ns the occupational radiation exposure NUREG 0386 006 R02: UNITED STATES NUCLEAR REGULA-information that has been reporWJ to the NRC's kadiation Ex.

TORY COMMISSION STAFF PRACTICE AND PROCEDURE posure Information Reporting System (REIRS).y nuclear power DIGcST Commission. Appeal Board Anet Licensing Board facihties and certain other categones of NRC hcensees dunng Decissons. July 1972 - June 1991

• Othce of the Generai Ccun-the years 1969 through 1989. The bulk of the data presented in sel (Post 860701). May 1992. 696pp 9206150288. 62024:001.

the report was obtained from annual radiation exposure reports This 2nd revision of the sixth edition of the NRC Practice and submitted in accordance with the requirements of 10 CFR Procedure Digest contains a dejest of a number of Commission, 20 407 and the technical specifications of nticlear power plants.

Atomtc Safety and Licensing Appeal Board, and Atomic Safety Data on worliers terminahng their employment at certain NRC and Licen9ng Board decisions issued dunng the penod of July bcensed facihties were obtained from reports submitted pursu-1,1972 to June 30,1991, interpreting the NRC's Rules of Prac-ant to 10 CFR 20 408. The 1989 annual reports submitted by Ice in 13 CFR Part 2.

about 448 bcensees indicated that approximately 216.294 indi-1

-i 2

Main Litations and Abstracts

- vduals were monitored. 111,000 of whom were morutored by Network.11 presents the radiation levels measured in the vicinity nuclear powor facilities. They L.tred an average individual of NRC licensed facihtes throughout the country for the first dose of 018 rem (cSv) and an average measurable dose of quarter of 1992.

0.36 (cSv). Termination radiation exposure reports were ana-ly7ed to reveal that about 113,535 individuals completed their NUREG-0847 SO9: SAFETY EVALUATION REPORT RELATED employment with one or more of the 448 covered hcensees TO THE OPERATION OF WATTS BAR NUCLEAR dunrg 1989. Some 76,561 of these individuals were terminated PLANT, UNITS 1 AND 2. Docket Nos. 50 390 And 50 391.(Ten-from power reactor facahteos, and about 10,344 of them were nesse Valley Authonty) TAM,P.S Dwision of Reactor Prejects -

considered to be transient workers who escowed an average 1/11 (Post 870411) June 1992 153pp. 9207140239. 62372.327.

dose of 0 64 rem (cSv).

Supplement No. 9 to the Safety Evaluation Report for the ap-NUREG 0725 R08: PUBUC INFORMATION CIRCULAR FOR plication filed by the Tennessee Valley Authonty for license to SHIPMENTS OF 1RRADIATED REACTOR FUEL

• D# vision of operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos.

Safeguards & Transportation (Post 870413). June 1992. 3Spp.

50 390 and 50-391, located in Rhea County. Tennessee, has 9207140034 62373.189.

been prepared by the Office of Nuclear Reactor Regulation of This circular has ben prepared to provide information on the the Nuclear Regulatory Commission, The purpose of this sup-shipment of irra6ated reactor fuel (spent fuel) subject to regula.

plement is to update the Safety Evaluation of (1) additional in-tion by the Nuclear Regulatory Commission (NRC), and to meet formation submitted by the apphcant since Supplement No. 8 the roquiremonts of Pubhc Law 96-295 The repor1 provides a was issued, and (2) rnatters that the statt had under review brief description of NRC authority for certain aspects of trans.

when Supplement No 8 was issued.

porting spent fuel. It provides desenptwo statist:cs on spent fuel NUREG-0915 R02: NRC COMPREHENSIV! cCORDS DISPOSl<

shipments regulatad by the NRC from 1979 to 1991, it also lists detailed highway and railway segments used within each stato TlON SCHEDULE.

• Division of #.. non Support Services from Octrber 1, t987 through December 31,1991, (Post 890205) March 1992. 341pp. 9204210086 61390:236 -

• "" "9 ""

hvREG 0750 V34102: INDEXES TO NUCLEAR REGULATOR'Y regulatons cited in the' General Services Administra.

monts, COMMISSION ISSUANCES. July December 1991.

• Dwision of ton's (GSA) " Federal information Resources Management Reg-Freedom of Information & Pubhcatens Services (Post 890205).

ulations" (FIRMR), Part 2019, " Creation, Maintenance, and Use Apol 1992. 50op. 9205150000. 61693.035.

of Records," and regulations issued by the Natonal Archives Digests and indexes for issuances of the Commisson, the and Records Administration (NARA) in 36 CFR Chapter Xll, Atomic Safety and Licensing Board Panel, the Adm,nistratwe Subchapter B, " Records Management," require each agency to law Judqes, the Directors' Decisions, and the Dentals of Pots' prepare and issue a comprehenswe records disposition sched-tions for Rulemaking are presentM ule that contains the NARA approved records disposition sched-NUREG-0750 V35101: INDEXES TO NUCLEAR REGULATORY ules for records unique to the agency and contains tha WJtNs COMMISSION ISSUANCES January-Much 1992.

• Division of General Records Schedules for records common to several or Freedom of information & Pubhcatons Services (Post 890205).

all agencies. The approved records dispo:,. tion schedules speci.

June 1992. 38pp. 9207160275. 62404 269 fy the appropriato duration of retention and the final disposition Cee NUREG-0750,V34.102 abstract.

for records created or maintarned by the NRC. NUREG-0910, Rev. 2, conta ns "NRC's Comprehenswe Records Disposition NUREG 0750 V35 NO2: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR FEBRUARY 1992. Pages 47-82.

• Dwision of Schedule," and the onginal authori?cd approved Citaten num-Freedom of Information & Pubhcatens Services (Post 890205).

bors issued by NARA. Rev. 2 totally reorganizes the records Apnl 1992. 42pp. 9205150078. 61696:019.

schodules from a functional arrangement to an arrangement by Legal issuancos of the Commission, the Atomic Safety and Li, the host office. A subject index and a conversion table have censing Board Panol, the AdministratNe Law Judges, and NRC also boon developed for the NRC schedulos to allow staff to Program Ottices are presented.

identify the r't.* achedule numbers easily and to improve their abihty to locate applicable schedules.

NUREG-0750 V35 NO3: NUCLEAR REGULATORY COMMISSION ISSUANCES FOR MARCH 1992 Pages 83144.

• Dwision of NUREG-0936 V11 N01: NRC REGULATORY AGCNDA.Ouarterly Freedom of 'nformaton & Publications Services (Post 890205).

Report. January-March 1992.

• Duision of Freedom of informa-May 1992. I Opp. 9206120312. 61990:023.

tron & Pubhcations Services (Post 890205). May 1992.140pp.

See NUREG-0750,V35.N02 abstract 9206030197. 61873:045.

NUREG-0750 V35 N04: NUCLEAR REGULATORY COMMISSION The NRC Regulatory Agenda is a compilation of all rules on ISSUANCES FOR APRIL 1992 Pago 145187.

• Dwision of which the NRC has recently completed action, or has proposed Freedom of Information & Publications Services (Post 890205).

action, or is considenng action, and all petitions for rulemalung June 1992. 40pp. 9206300240. 62183:220.

which have been received by the Commission and are perdng See NUREG 0750.V35,N02 abstract.

disposition by the Commisson. The Regulatory Agenda is up-.

dated and issued each quarter.

NUREG 0837 V11 N04: NRC TLD DIRECT RADIATION MONI.

TORING NETWORK Progress Report. October-December.1991.

NUREG-0940 V11 NOI: ENFORCEMENT ACTIONS. SIGNIFI-STRUCKMEVER,R4 MCNAMARA,N. Region 1 (Post 820201)-

CANT

- ACTIONS RESOLVED. Quarterly Progress April 1992. 328pp, 9204210093. 61392:001-Report,Januar, M rch 1992.

Ofc of Enforcement (Post This report provides the status and results of the NRC Ther-870413). May 1992. 407pp 9206080153. 61917:007.

moluminoscent Oosameter (TLD) Direct Radiaton Monitonng This compilation summanzes significant enforcement actions Network, it presents the radiation levels measured in the vicinity that have been resolved dunng one quarterly penod (January +

of NRC licensed facilities throughout the country for the fourth March 1992) and includes copies of letters, Notices, and Orders

@aw of M sent by the Nuclear Regulatory Comrnission to licensees with NUREG-0837 V12 N01: NRC TLD DIRECT RADIATION MONb respect to those enforcement actions It is anticipated that the TORING NETWORK. Progress Report. January-March,1992.

information in this pubhcation will be widely disseminated to f

STRUCKMEYER,Ra MCNAMARA,N Regen 1 (Post 820201) managers and employees engaged in actwities licensed by the June 1992. 234pp. 9207020042. 62188:103 NRC, so that actons can be taken to improve safety by avoid-l Thrs report provides the status and results of the NRC Ther-ing future violatons similar to those described in this publica-moluminescent Dosimeter (TLD) Direct Radiation Monitanng tion.

i r

,y _

Main Citations and Abstracts 3

HUREG 1125 V13. A CCMPILATION OF REPORTS OF THE AD-tion is reviewed by the NRC for consistency only and no ince-VISORY COMM:TTE E ON REACTOR SAF EGUARDS 1991 pendent vahdahon and/or venficabon is pedormed. For detailed Annual

• ACHS - Adesory Committee on Nactor Safeguards.

und complete informat6on about tables and figures, refer to the Apnl 1992 139ep 9205140216 61669 001.

source pubhcations This digest is pubbshed annually.

This comt..lation contains 41 ACRS reports submitted to the NUREG 1414: DlFF ERING PROFESSIONAL VIEWS OR OPIN-Commisson, the Execuhvo Director for Operat ons, or to th, Office of Nuclear Regulatory Research, donng calendar yea, IONS 1990 Special Review Panel

• NRC - No Detailed Affitr anon Given. Apnl 1992. 70pp 9205140213 61668159 1991. it also includes a report to Congress on the NRC Safety in December 1989 the Evocutive Director Operations of the Research Protyam All reporis have been made available to the pubhc through the NRC Pubhc Document Room and the U S Li-U S Nuclear Regulatory Commission (NRC) appointed a Special Amw Panel to mam N eHecWomss of E Manual brary of Congross. The reports are deded into two groups Part

1. ACRS Reports on Project Reviews, and Part 2, ACRS Re.

Chaptcir 4125, Diftenng Professional Views or Opinions, und ports on Genenc Subrects. Part 1 contains ACRS mports alpha.

NRC Manual Chapter 4126, Open Door Pobcy. In accordance wth Section E of NRC Appendiv 4125, the Panel was responsi-belge' ay pro ict name and by chronological ordor within protect name. Part 2 categonzes the reports by the most appro, ble for assessing ' the informal and formal processes for deahng with d ffonng professional views or opinions, including pnate Genenc sublact area and by chronological order within the effectiveuss of the processes, how well they are under-sublect area stood by employees, and the organizahonal climate for having NUREG-1266 V06: NRC SAFETY RESEARCH IN SUPPORT OF these views and opinions aired and property decidnd." This REGULATION FY 1991

• Othce of Nuclear Regulatory Re-report preserits the Special Review Panel's evaluation of the search (Post 660720) Apn! 1992 88pt.

9205200069 NRC s cun,ni arr ss for dealing with Differing Professional 61734 269 Views c % - M rovided in this report are the results of an This report, the seventh in a senes of annual reports, was empioyee opr on survey on the process; high!ights and sugges-prepared in response to congresuonal inquines concerning how t ons from interviews with individuals who had submitted a Dif.

nuclear regulatory research is used it summanzes the eccom-fenng Professional View or Opinion, as well as with agency phshments of the Office of Nuclear Regulatory Research dunn9 managen directly involved with the Diffonng Professional Views FY Ib91. The goal of this office is to ensure that safety-related or Opinines process, and proposed revisio-s to Manual Chap-research provides the technical bases for rulemaking and for re-ters 4125 and 4126.

lated decisions in support of NRC licensing and inspection ac-tivities T h s research is necessary to make certain that the reg.

NUREG 1410 V04 NO2: OFFICE OF THE INSPECTOR utations that are impos9d on Lcensees provide an adequate GENERAL Semiannual Report. October 1,

1991 March margin of safety so as to protect the health and safety of the 31,1991

• Office of the inspector General (Post 890417) Apnl pubhc.1Ns report describos both the direct contnbuSons to sce 199141pp. 9206120319 62022 035.

entific and technical knowledge with regard to nuclear safety inspectors Generals are required, by the IG Act of 1978. as and thmr regulatory apphcations.

amended. to prepare semiannual reports which summanze the significant investigative and audit activites of the office. The 6-HUREG 1327: INITIAL DEMONSTRATION OF THE U S. NRC'S montn reporting penod ends Much 31 and September 30. The CAPADILitY TO CONOUCT A PERFORMANCE ASSESSMENT report is submitted tn *

' airman not later than April 30 and FOR A HIGH LEVEL WASTE REPOSITORY CODELL.Ra October 31, respecta J each yex. The Chairman prepares EMENBERG.N; FEHHiNGER.DJ et al. Office of Nucicar Maton-comments as requ' the IG Act, and transmits the report to al Safety & Safeguards. May 1992. 167pp. 9206150272.

Congress 62022.125 In order to bottor review licensing submittals for a High-Level NUREG-1419: DIRECTORY OF CERTIFICATES OF COMPLl-Wasto Repository, the U S Nuclear Regulatory Commission ANCE FOR DRY SPENT FUEL STORAGE CASKS.

• Dmsson of staff has onpanded and improved its capabihty to conduct per.

Industnal & Medical Nuclear Safety (Post 870729). February formance assessments This report documents an instal demon-1992. 53pp D205160143 61695:222.

stration of this capabihty. The demonstrat:on made uso of the This directory contains: (1) Certificates of Compliance for all hmited data from Yucca Mountain. Nevada to investigate a dry spunt fuel storage casks approved by the U S. Nuclear Reg-small set of scenano classes Models of release and transport utatory Commission, and (2) Summary Reports of each ap-of radionuclides from a repository via the groundwater and proved cask model Later directones wtil contain. (1) a list of direct release pathways provided prehmanary estimates of re-cask users, and (2) a hst of cask locations. The purpose of this leases to the accessible environment for a 10.000 year simula.

directory is to make avMable a convenient source of informa-tion tima Lat n hypercube sampung of input parameters was tion on spent fuel storage casks which have been approved by used to express results as distobutions and to investigate modet the U S. Nuclear Regulatory Commission. Storage of fuel as-sensilmbes This methodology demonstrabon should not Le in-sembhes umng these casks must bo in accordance with the pro-torpreted as an estimate of performance of the proposed repos-visions of 10 CFR Part 72 Itory at Yucen Mcuntain. Nevada' NUREG-1422:

SUMMARY

OF CHERNOBYL FOLLOWUP RE-NUREG 1350 V04; NUCLE AR REGULATORY COMMISSION IN-SEARCH ACilVlilES

• Dmsion of Safety issue Re$olution FORMAllON DIGEST 1992 Edition. OUVEKL Dmsson of (Post B80717) June 1992. 74pp. 9206250386. 62112 273.

Budget. March 1992.130pp 9206120335 61998 092.

In NUREG-1251 "Imphcations of the Accident at Chernobyl The Nuclear Regulatory Commission information Digest pro-for Safety Regulation of Commercial Nuclear Power Plants in vides a summa y of information about the U S Nuclear Regula-the United States." Apnt 1980, the staff of the U S Nuclear tory Commission (NRC). NRC s regulatory responsibihties, the Regulatory Commission (NRC) concluded that no immediate act.vities NRC hcenses. and general informabon on domest:c changes in NRC s regulabons regarding design or operahon of and worldwido nuclear energy Tnis digest is a compilation of U S. commercial reactors were noeced, however, it recommend-nuclear-and NRC-related data and is demgned to provide a ed that certain issues be consdered further. NRC's Chemobyt quick roMrence to major facts about the agency and the indus.

followup research program consisted of the research tasks un-try it requtates In genmal the data cover 1975 through 1991, dertaken in response to the recommendations in NUREG-1251.

with except:ons noted Information on generatmg capacity and It included 23 tasks that addressed potential lessons to be average capsicity factor for operahng U S commercia! nucicar learned from the Chernobyl acctdont. This report presents sum 4 rower reactors is obtained from monthly operating reports that manes of NR~s Chernobyl foHowup research tasks. For each arn submitted dire-tty to the NRC by the hcenue This informa-task, the CherncOyLrelated 65ues am indicated. the work is de-

4 Main Citations and Abstracts scnbod, and the staff s findings and conclusions are presented da, Maryiand. during the week of October 20-30,1991. The More dotaded reports concerning We work are referenced papers are pnnted in tne ordor of their presentation in each ses-where apphcable. Tb s report closes out NRC's Chernobyl fol.

siun ano desenbe progress and results of programs in nuclear lowup research program as such, but ad3tional research will be safety research conducted in this country and aDroad Foregn conducted on some issues as needed The report tncludes re-participation in the moeting included 14 different papers pre-marks concerning significant further activity witn respect to the sented by researchers from Canada, Germany, France, Japan.

issues addressed.

Sweden, Taiwan and USSR. The tities of the papers and NUREG 1447: STANDARD REVIEW PLAN UPDATE AND DEVEL-names of the authors have been updated and may differ from OPMENT PROGRAM IMPLEMENTING PROCEDURES DOCU_

those that appeared in the final program of the meeting MENT

• Program Management, Pohey Development & Ana!ysis NUREG/CP-0119 V02: PROCEEDINGS OF THE NINETEENTH Staff (Post 870411). May 1932. 239pp 9206300250. 62187.224 WATER REACTOR SAFETY INFORMATION MEETING.

This implementrng procadures document (IPO) was prepared WEISS.Al Brookhaven National Laboratory. April 1992.489pp.

for use in implementing tasks undta the standard review plan 9206080161. 61930'001.

update and development program O P-UDP) The IPD provides See NUREG/CP-0119,V01 abstract.

comprehensive guidance and detaied procedures for SRP UDP l

tasks The IPO is mandatory for contractors performing work for the SAP-UDP. It is guidance for the staff. At the completion of WATER REACTOR SAFETY !NFORMATION MEETING.

the SRP.UDP, the IPO will be revised (to rerr ove the UDP as.

WEISS, A.J. Brookhaven National Laboratory. Apne in92. ?80p.

92 8 6 3 0

pects) and will replace NRR Of6ce Letter No. 800 as long term maintenance procedures.

p 9

NUREG-1452; REVIEW AND EVALUATION OF NUREGiCP-0119 V03 AD: PROCEEDINGS OF THE NINE-TECHNOLOGY,EOUtPMENT, CODES AND STANDARDS FOR 1EENTH WATER REACTOR SAFETY INFORMAT!ON MEET-O!GITIZATION OF INDUSTRIAL RADIOGRAPHIC FiL M ING, WEISS A.J. Brookhaven National Laboratory. June 1992.

MUSCARA,J Task Group on Digitization of Industnal Radio.

24pp 9207140232. 62373122, graphs May 1992. 41pp 9206120316. 61999 321 See NUREG/CP 01?9,V01 abstract.

This report contains a review and evaluation of the technolo-NUREG/CR-2000 VII N2: LICENSEE EVENT REPORT (LER) gy, equipment, and codes and standards related to the digitaa' COMPILATION For Month Of February 1992.

• Oak Ridge Na-tion of industna! radographic fdm. The report presents recom-tiona, Laboratory. March 1992.144pp. 9204210068 ORNL/

mendations and equipment performance specifications that v.ill NSIC-200. 61393 001, allow the digitaabon of radiographic film from nuclear power This monthly report contains Licensee Event Report (LER) plant components in order to produce faithful reproductions of operat ona! information that was processed into the LER data flaw images of interest on the films Just.fication for the speci4 fde of the Nuclear Operations Analysts Center (NOAC) dunng cations selected are provided Performance demonstration tests the one month penod identified on the cover of the document.

for the digit:zation process a e required and cntena for such The LERs, from which this information is denved, are submitted tests is presented Also severW comments related to implemen' to the Nuclev Regulatory Commission (NRC) by nuclear power tatJon of the technoingy are presented and discussed plant licensees in accordance with federal regulations. Proce-AUREG 1456: AN ALTERNATIVE FORMAT FOR CATEGORY I dures for LER reporting for reywons to those events occumng FUEL CYCLE FACILITY PHYSICAL PROTECTION PLANS poor to 1984 are described in NRC Regulatery Guide 1,16 snd DWY ER.P. A.

Division of Safeguards & Transportaten (Post NUREG 0161, "tnstructions for Preparation of Data Entry 870413). June 1992. 2Bpp 9207020033. 62100 001.

Sheets for Licersee Event Reports." For those events occuning Tnis document provides an alternative format for phys;.at pro.

On and after January 1.1964, LERs are being submitted in ac-tection plans designed to meet the requirements of Title 10 of cordance with me revised rule contained in Titk 10 Part 50,73 the Code of Federa! Regulations. Sections 73 20, 73 45, and of the Code of Federal Regulations (10 CFR 50.73 Licensee 73 46 These requirements apply to licensees who operate Cat-Event Report System) which was pubbshed ;n the Federal Reg-egory I fuel cycle f acilities Such kcensees are authonzed to use ister (Vol 48, No 144) on July 26,1983. NUREG-1022, "Li-or possess a formula quantity of strategic pecial nuclear mate-censee Event Report System.

Desenption of Systems and r:al The format descnbed is an attemative to that found under Guidehnes for Reportmg." proudes supporting guidance and in-Regulatory Guide 5 52, Rev. 2. " Standard Format and Content formation on the revised LER rule The LER summanes in this of a Licensee Physical Protecnon Plan for Strategic Special Nu.

report are arranged alphabetically by facihty name and then clear Material at Fixed Sites (Other than Nucity Power chronologicany by event da:e for each facihty.- Component, Plantsr system, keyword, and component vendor indexes follow the NUREG-1458: EMERGENCY RESPONSE TO A HIGHWAY ACCI-DENT IN SPRINGFIELD, MASSACHUSETTS.ON DECEMBER 9

16.1991.

Divtson of Safeguards & Transportaten (Post 870413) June 1992.128pp. 9207140291,62371'001.

S to '

On December 16.1991, a truck carrying unirradiated (fresh) nuclear fuel was irivolved in an accident on U S. Interstate 91, NUREG/CR 2000 V11 N3: LICENSEE EVENT REPORT (LER) in Spnngfield, Massachusetts This report descnbes tne emer.

COMPILATION For Month Of March 1992.

• Oak Ridge Nation-gency response rneasures uncertaken by local State, Federal al Laboratory. May 1992 61pp. 9206120288 ORNL/NSIC-200.

and povate parties. The report also discusses " lessons 6 t 999.262.

learned" from the response to the accident and suggests areas See NUREG/CR-2000,v11,N02 abstract.

where improvements might be made.

NUREG/CH-4219 V08 N2: HEAVY SECTION STEEL TECHNOL-NUREG/CP-0119 VO1: PROCEEDINGS OF THE NINETEENTH OGy PROGRAM. Semiannual Progress Report For April-Sep-WATER REACTOR SAFETY INFORMATION MEE TING.

tember 1991. PENNELL.1l.E. Oak Rdge National Laboratory.

WEISS A.J Brookhaven Natonal Laboratory. Arnt 1992 514pp Apnl1992.102pp.9205150011. ORNL/TM-9593. 61695:277, 9206080159. 61918 134.

The Heavy-Section Steel Technology (HSST) Program is,,on-This three-volume report conta:ns 83 papers out of the 108 ducted for the Nuclear Regulatory Commission by Oak Ridge that were presented at the Nineteenth Water Reactor Safety in-Nationat Laboratory (ORNL) The program focus is on the de-formaton Meeting held St the Bethe3da Mamott Hotel Bethe&-

Velopment and vabdahon of technology for the assessment of

Main Citations and Abstracts 5

fracture prevention margins in rommercial nuclear reacto pres-and wil! extend for 4 years The intent of the program is to sure vessels. The HSST Program is organized in 10 tasxs- (1) venfy and improve fracture analyses for circumferentially program management, (2) fracture methodology and analysis, cracked largo-diameter nuclear piping with crack sees typicalty (3) matonal characterization and p<opertes. (4) Spec:al technical used in leak before-break analyses or inaervice flaw evalua-assistance, (S) fracture anatysis computer programs, (6) cleav-tions. Only qu;si-static loajing rates are esaluated since the ago-crack trutiatiort (/) cladding evaluations, (8) pressunze/

NRC's International Piping Integnty Research Group (IPIRG) thermal shock technology, (9) analysni methode vahdation, and program is evaluaimg the effects of seismic loadirj rates on (10) fracture evaluation tests The progra.n tasks have been cracked piping systems Progress for through wall-cracked pipe structured to place emphasis on the fesolut:on of fracture involved. (1) conducting a 28 inch diameter stainless steel SAW issues wrth near-term licensing significance Resources to exe-and 4 inch diameter French TP316 expenments, (2) conducting cute the research tasks are drawn from 09NL with subcontract a matnx of FEM analyses to determine GE/EPHI functions for support from universities and other research laboratones. Close short TWC pipe, (3) companson of uncracked pipe maximum contaci is maintained with related research programs both in moments to vanous analyses and FEM solutions, and (4) devel-the United States and abroad.

opment of a J-estimation scheme that includes the strength of both the weld and base metals. Progress for surface-cracked NUREG/CR-4469 V12: NONDESTRUCTIVE EXAMINATION (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT pipe involved. (1) conducting two experiments on 6-inch diame-ter (Sch. 40 and XXS) pipe with d/t = 0.5 and 0/t1 - 0.25 WATER REACTORS. Annual Report, October 1989 September cracks, (2) compansons of the pipe expenments to Net-Section-1990. DOCTOR.S.RJ GOOD.M S.; HEASLER,P Ga et at Bat-Col apse predictions, and (3) modification of the SC TNP and tolle Memonal Institute, Pacific Northwest Laboratory. May SC.TKP J-estimation schemes to include extemal surface 1992. 40pp. 9206150280. PNL 5711. 62022:086.

The Evaluation and improvement of NDE Reliability for in-cracks. High-temperature hardness testing appears to be a j

useful screening entona parameter for assessing the susceptibil-serwce inspection of Light Water Reactors (NDE Reliabihty) ity of ferritic pipe to dynamic strain aging For anisotropic frro Program at the Pacific Northwest Laboratory was established by ture ovaluations, it was found that only one of five femtic pipes the Nuclear Regulatory Comnussion to determine the reliabihty had the low toughness direction in a helical direction, the rest of current inserwce inspection (ISI) techniques and to develop had low toughness in the axial d:rection. For crack opening area recommendations that will ensure a suitably high inspection reli.

analyses, predictive capabihties were expanded so that load abihty. The obloctives of this program include: determining the versus crack opening can be calculated from the LBB.NRC, reliability of ISI performed on the primary systems; of commer-GE/EPRI, LBB.GE. LBB ENG, and Tada/ Pans analyses. These cial hght. water reactors (LWRs), using probabilistic fracture me-chanics analysis to determine the impact of NDE unrehabihty on include i ading due to tension, bending, and combined tension aM Wg W WTM anaWS was a% WM b ao system safety; and evaluahng reliabihty improvements that can count for the weld and base metals strengths. t:lastic FEA be achieved with improvm1 and advanced technology. A final sh wed that for pressure loading. a crack close to a terminal ot,ectNe is to formulate recommended revis6cns to ASME Code end (i e.. a nozzio) will have lower crack opening due to re-and Regulatory requirements, based on matenal proporties, straint of the induced bending This could affect LBB analyses.

service conditions and NDE uncertainties The program scope is lim.ted to tSi of the pnmary systems including the piping, NUREG/CR 4753 V05:

CANADIAN SEISMIC vessel. and other components inspected in acce"hnce w!!h AGREEMENT.Annuat Report, July 1989 June 1990.

Section XI of the ASME Code. This is a progress report cover

• WETMILLER,R.Ja LYONS,J.Aa SHANNON.W.Ea et al, Canadi-ing the programmatic work from October 1989 through Septem' an Commercial Corp. May 1992. 49pp. 9205270006. 61769-178.

bw 19M This is e progress report of work camed out under the terms NUREG/CR 4551 V2R1P4: EVALUATION OF SEVERE ACCI.

of a research agreement entitled the " Canadian Seismic Agree-DENT RISKS: OUANTIFICATION OF MAJOR INPUT ment" between the U.S Nuclear Regulatory Commission PARAMETERS Experts' Determination Of Source Term Issues.

(USNRC), the Canadian Commercial Corporation and the Geo-HARPER,F.T.: BREEDING.R J.; BROWN T.D ; et a!. Sandit Na.

physics Division of the Geological Survey of Canada (GD/GSC) tional Laboratones. June 1992. 502pp. 9207060085. SAND 86-du'ing the penod from July 01, 1989 to June 30,1990. The 1309 02195 250.

" Canadian Seismic Agreement" supports generally the oper.

In support of the Nuclear Regulatory Commission's (NRC's) ation of various seismograph stations in eastern Canada and assessment of the nsk from severe accidents at commercial nu-the collection and analysis of earthquake data for the purpose clear power plants in tne U S reported in NLMG-1150, the of mitigating seismic hazards in eastem Canada and the north-Severe Accident Risk Reduchon Program (SARRP) has com.

eastem U.S The specific activities camed out in this one year pleted a revtsed calculation of the nsk to the general pubhc penod are summanzed below under four headings: Eastem from severe accidents at five nuclear power plants: Surry, Se.

Canada Telemetred Network and local network developments, quoyah. Zion. Peach Bottom. and Grand Gulf. The emphasis in Dataiab developments; strong-motton network developments, this nsk analysis was not on determining a point estimate of and earthquake activity During this penod the first surface fault nsk, but to determine the distnbution of risk, and to assess the unequivocably determined to have accompanied a histonc uncertainties that account for the breadth of this distribution.

earthquake in eastern North Amenca, occurred in rorthern Off site risk initiation by events, both intema! to the power sta.

Quebec _

tion and extemal to the power station were assessed Much of NUREG/CR-4831: STATE OF THE ART IN EVACUATION TIME the important input to the logic models was generated by export ESTIMATE STUDIES FOR NUCLEAR POWER PLANTS, panels This document presents the distnbutions and the ration-URBAN 1K,T.E.; JAMISON.J.D. Battelle Memonal Institute, Paci!.

le pportin0 he distnbutions for the questions posed to the ic Northwest Laboratory March.992.19pp. 9204210103. PNL-7776. 61391:2 t 7.

NUREG/CR 4599 V01 N2: SHORT CRACKS IN P! PING AND in the event of a maior accident at a commercial nuclear PIPING WELDS Semiannual Report, October 1990 March power station, exposure of the pubhc to airbome radioactive 199t, WILKOWSKI.G M BRUST,Fa FRANCINLR.; et at Bat.

matena!s can be prevented or greatly reduced by evacuating telle Memonal Institute, Columbus Laboratories. April 1992 the area immediately surrounding the reactor site. Reactor h-202pp 9205200073 BMi-2173 61732.001.

censees are required to conduct studies to estimate the time This is the second semiannual report of the U S Nuciear needed to evacuate the public from the area surrounding each Regulatory Commiss+on's Short Cracks in Piping and Piping nuclear power station. The results of such atudies are used by Weld? research program The program began in March 1990 regulatory personnet and emergency panners to assess the po-

[

6 Main Citations and Abstracts

- tontial effectiveness of protective responses for the public. The Idaho National Engineering Laboratory. MAR D's pnmary func-time required to evacuate the pubhc from a 10 mile emergency tion L4 to create a data repository for NUREG-1150 and other

. planning radius is estimated by analyzing the available trarapor-permanent data by providing input, converson, and output ca-tation facihties and other relevant conditions within this radius.

pabihties for data used by IRRAS, SARA. SETS, and FRANTIC To support the analysis, data must be collected and assump-personal computer (PC) codes. As probabehstic nsk assess-tons must be made regarding the transportaten facilities, the monts and individual plant examinations are submitted to the size and charactor'stics of the population and other condt ons NRC for review, MAR D can be used to convert the models and i

in the planning zone Thss report dosenbes standard approach-results from the study for use with IRRAS and SARA. Then, es and provides recommendahons regarding the relevant infor-these data can be easily accessed by future studies and will be.

mation, assumptons and mothods to be used in performing in a form that will enhance the ana!ysis process. This reference evacuation time estimate stud es.

manual provides an overview of the functions availalle within NUREG/CR 5114: FINDINGS OF A WORKSHOP ON DEVELOP-the MAR-D and step-by-step operating instructons.

ING A METHODOLOGY FOR EVALUATING EFFECTIVENESS NUREG/CR 5445: PERFORMANCE OF INTACT AND PARTIAL.

OF NUCLEAR POWER PLANT TRAINING. CANTOR.J.A.,

LY DEGRADED CONCRETE BARRIERS IN LIMITING MASS FULLER,R E.; WALKER C.L.: et at DOL Omni Engineenng Corp TRANSPORT. WALTON J.C. EG&G Idaho, Inc. June 1992.

April 1992. 49pp 9205180160. 61710:186-39pp. 9207060042. EGG 2662. 62197:032.

In October 1990, the Nuclear Regulatory Commission (NRC)

Mass transport through concrete barners and release rate sponsored a workshop to develop a proposed methodology for from concrete vaults are quantitatively evaluated. The thorny use by the NRC in deterrnining the effectiveness of nuclear utili-tssue of appropriate diffusion coefficients for use in performance ty training The workshop developed a framework on which to assessment calculabons is covered, with no ultimate solution base a methodology which draws together cunent NRC and nu~

found Release from monohthic concreto vaults composed of clear industry processee and initiatives in training evaluation and ccncrete waste forms is esttmated with a semeanalytical solu-plant performence monitonng. The framework recognizes that tion. A parametric study illustrates the importance of different utihties, under cuirent NRC and industry guidance, operate parameters on release. A second situaton of importance is the closed loop training systems that incorporate methods foi self-role of a concrete sholl of vault placed around typical waste correenon. The model proposes that by monitonng/samphng in-forms in limiting mass transport. In both situations, the pnmary dicator data at vanous points in the utihty's closed-loop system, factor controlkng concrete performance is cracks. The imphca-the NRC can determino whether the loop is operating properly tions of loaching behavior on likely groundwater concentrahons to maintam training program effectiveness. This training loop in-is examined Frequently, lower groundwater concentrations can ciudes the training process, the performance of trained workers, be expected in the absence of engineered covers that reduce and plant operators Monitoring / sampling of indicators is infiltration.

planned such that each indicator provides data which Comple-ments data denved from the other ndicators. Agreement be-NUREG/CR 5569: HEALTH PHYSICS POSITIONS DATA BASE.

tween indicators is used to confirm either effective training or to KERR,G D.; BORGES,T.; STAFFORD,R.S.; et al. Oak Ridge Na-detect training problems. Inconsistency between indicators tng-tional Laboratory. May 1992. 254pp. 9206260350, ORNL/TM-gets further investigation.

12067,62121:333.

NUREG/CR 5121: EXPERIMENTAL RESULTS FROM PRES-This report is a collection of summanos of the Health Physics SURE TESTING A 1.6-SCALE tlUCLEAR POWER PLANT Positions (HPPOS) Data Dase of the Nuclear Regulatory Com-(

CONTAINMENT, HORSCHEL.D.S. Sandia National Laborato-mission (NRC) on a wide range of topics in radiation protection ries. January 1992. 2,035pp. 9204210107. SAND 884006 (health physics). The HPPOS data base consista of 247 ongsnal 61385:001.

documents in the form of letters, nemoranda, and excorpts This report discusses the testing of a 1:6-scale reinforced.

from technical reports. It was developed by the NRC Headquar.

concrete containment building at Sandia National Laboratones ters and Regional Offices to help ensure uniformity in inspec+

in Albuquerque, New Monico. The scale-model containment l'ons, enforcement, and licensing actions A vanety of indexing building was designed and built to the Amoncan Society of Me, schemes were used in this report to increase its usefulness.

chanical Eng neers coda by United Engineers and Constructors.

The summanes in this report are wntten in the context of the Inc., and was instrumented with over 1200 transducers to pre-current Part 20 of Title 10 to the Code of Federal Regulations pare for the test Tho containment model was tested to failure (10 CFR Part 20 Sections 20.1 20.601). TM HPPOS summa-to determine its response to static internal overpressunzation.

ries and onginal documents are intended to serve as a source As part of the U S. Nuclear Regulatory Commission's program of mformation tnat will be useful in radiation protection pro-on containment integnty, the test results wi., be used to easess grams at nuclear research and power reactors, nuclear medical the capatxhty of analytscal methods to predict the performance laboratones and other industnes that either process or use nu-of containments under severe accident loads. The scaled ge.

clear matenals. The purpose of this report is to allow interested ometry of the model was typical of a full-sce containment The individuals to famihanze themselves with the contents of the containment inctoded equipment hatches, personnel air locks, HPPOS Data Base and with the basis of many NRC decisions several small piping penetrations, and a thin steel liner that was and regulations.

attached to the concrete by headed studs. The containment NUREG/CR-5604 VOI: ASSESSMENT OF ISLOCA RISK METH/

model and its instrumentation are bnefty discussed and is foi-OOOLOGY AND APPLICATION TO A BABCOCK AND WILCOX lowed by the testing procedures and rneasured response of the NUCLEAR POWER PLANT. Main Report. GALYEAN.W.J.;

containment model. A summary is included to aid in understand-i GERTMAN,D.L EG8G idaho, Inc. AprW 1992. 79pp.

ing the test as it apphes to real-world reinforced concrete con-9205270010 EGG 2600. 61769 220 taenment structures. The data gathered dunng testing are includ-This document presents information essential to understand-ed as an appendix ing the nsk associated with inter-system loss of coolant acci-NUREG/CR 5301: MODELS AND RESULTS DATABASE (MAR-dents (ISLOCAs) The methodology developod and presented in D), VERSION 4.0. Reference Manual. BRANHAM-HAAR K; this document provides a state-of the. art method for identdying DINNEEN,R.A.; RUSSEL LK D,; et al. EG&G Idaho, Inc May and evaluating plant-specific hardware designs, human perform-l-

1992.143pp 9206250293 EGG 2627. 02110001.

ance issues. and accident consequence factors relevant to the The Nuclear Regulatory Commission's Office of Nuclear Reg-prediction of the ISLOCA nsk. This ISLOCA methodology was ufatory Research (NRC.RES) is presently funding the develop-developed and then apphed to a Babcock and Wilcox (B&W) ment of the Models and Results Database (MAR-D) at the nuclear power plant. The results from this apphcation are de-

~,.

,, +,

m l

Main Citations and Abstracts 7

scnbed in detait For this particular B&W reference piant. the as.

NUREG/CRs 33: AGING ASSESSMENT OF COMPONENT sessment inchcated that the probability of a severe ISLOCA is COOLING WATER SYSTEMS IN PHESSURIZED WATER RE-approximately 2.2E - 06/ reactor year.

ACTORS LOFARO,R,; GUNTHER.W.; SUDVDHl.M.; et al Brookhaven National Laboratory. June 1992.

50pp.

NUREG/CR 5604 V02: ASSESSMENT OF ISLOCA RISK METH-9207060066. ENL-NUREG-52283. 62197:137.

ODOLOGY AND APPLCATION TO A BABCOCK AND WILCOX A two-phase aging ana!ysis of component cooling water NUCt EAR POWER PLANT.Appendmes A-H GALYEAN.W.JJ (CCW) systems in pressurced water reactors ;P VRs) was per-GE RTM AN.D 1.

EG&G Idaho, Inc.

Apnl 1932 474pp formed. In phase I, wtuch was published separately as NUREG/

9205270067. EGG-2608. 61760 064.

CH-5052, the effects of aging on the CCW system were charac-See NUREG/CR 5604,V01 abstract.

tenzod, and the predominant f ailure modes aging mechanisms, and compononts susceptible to aging degradation were idenb-NUREG/CR 5604 V03: ASSESSMENT OF ISLOCA RISK - METH' fied Failure rate trends were examined, and their effect on ODOLOGY AND APPLICATION TO A BABCOCK AND WILCOX time dependent system unavailabdity was investigated in phase 11, which is the sub ect of this report, the methods used to NUCLE AR POWER PLANT. Appendices 1 M. GALYEAN,W J,

i GERTMAN D i EG&G Idaho, Inc Apol 1992 381 p' manage ag<ng degradation in the CCW system were studiert. In-9205270071. EGG-2008 61767:001-formation was collected and analyred on inspection, surveil-See NUREGICR 5604.V01 abstract lance, monitonng, and maintenance techniques from a survey of operating PWRs. The results are presented horein Advanced NUREG/CR 5633; AUXILIARY FEEDWATER SYSTEM RISK-techniques that may help detect and mityate aging degradstion BASED INSPECTION GUIDE FOR THE TURKEY POINT NU.

also are included. A detmied examination of the matenals of CLEAR POWER PLANT. MOF FITT,N E.; GORE,B.F.; VO.T V, construction and their relationship to vanous aging mechanisms Battelle Memonal institute, Pacific Northwest Laboratory Apnt is discussed in addition, the vanous standards, codes, and reg-1992. 33pp 9205140204 PNL-7454. 616F8 093 ulatory requirements that govern the design, construction, and in a study sponsored by the U S. Nuclear Regulatory Com-operation of the CCW system were investigated. Recommenda-mission (NRC), Pacific Northwest Laboratory has developed and tions to better manage aging degradation in component cooling apphed a methodology for derMng plant specihc nsk-based in-water system are discussed.

Spection guidance for the adwliary feedwater (AFW) system at pressunted water reactors that have not undergone probabilistic NUREG/CR 5710: STRESS-CORROSION CRACKING STUDIES nsk assessment (PRA) This methodology uses existing PRA re-ON CANDIDATE CONTAINER ALLOYS FOR THE TUFF RE-suhs and plant ope <ating expenence information. Existing PRA-POSITORY. BEAVERS.J A ; DURR.CL Cortest Columbus based inspection guidance information recently developed,for Technologies, Inc. (formerly Cortest Columbus, Inc ). May 1092.

the NRC for vanous plants was used to identify genenc compo-124pp. 9206250359. 62112.149.

nent failure modes. This information was then combined with Cortest Columbus Technologies, inc_ (CC Technologies) in-plant specthe and industry-wide component information and f ail-vestigated the long term performance of container matenals ure data to identify failure modes and failure mechanisms for used for high4evel wasto packages as part of the information the AFW system at the selected plants. Turkey Point Nuclear needed by the Nuclear Regulatory Commission (NRC) to assess Power Plant was selected as one of a senes of plants for study the Department of Energy's application to construct a geologic The product of this effort is a pnontued liating of AFW fa: lures repository for high-levet radioactive waste. At the direction of which have occurred at tL piant and at other PWRs. This list-the NRC the program focused on the Tutt Repository This ing is intended for use by NRC inspectors in the preparation of report summan7es the results of Stress. Corrosion-Cracking inspectron plans addressing AFW nsk-important compononts at (SCC) studies performed in Tasks 3, 5 and 7 of the program.

the Turkey Point plant Two test techniques were used U bond exposures and Slow-Strain Rate (SSR) tests. The testing was performed on two NUREG/CR 5672 V02; CHARACTERISTICS OF LOW-LEVEL RA-copper base alloys (Alloy CDA 102 and Alloy CDA 715) arJ two DIOACTIVE WASTE. Decontaminahon Waste Prugram, Annual Fe-Cr-Ni a!!cys (Alloy 304L and Allov 825) in simulated J-13 Repot ' ' 3r Fiscal Year 1991= MORCOS,N; MCCONNELL,J W4 groundwater and other simulated solutions for the Tuff Repost AKERwd W. EG&G Idaho. Inc. June 1992.58pp 9207060049-tory. These solutions were designed to simulate the effects of EGG-2635, 62197:075 concentration and irradiation on the groundwater compostiort The objective of the Low-Level Radioactive Waste--Decon.

Possible radiolysis products evaluated included H(2)O(2) and tamination Waste Program (FIN A6359), funded by the United NANO (2). All SCC testing on the Fe-Cr.Ni Alloys was performed States Nuclear Regulatory Comm:ssinn (NRC), is to provide on solution-annealed specimens and thus issues such as the base 4ine data on the physical stability and teachabihty of solid" offect of sensitization on SCC were not addressed All four fled wasto streams generated in the decontamination process alloys were resistant to SCC in the J 13 well water and in the J-of pnmary coolant systems in operatmg nuclear power stations-13 well water that was concentrated by a factor of about 80 by The data include information on the chemical composition and evaporation, Alloy 825 also was reestant to SCC in all other en-characten2ation of those waste streams In addition this work is vironments ovaluated including chionde solutions containing up intended to evaluate waste form charactenzation tests ident fied to 100 000 ppm CI(-) and H(2)O(2) On the other hand, Alicy in the " Technical Position on Waste Form" (Revision 1), pro-304L underwent SCC in several of the CI(-) containing solutions.

pared by the Low Level Waste Management Branch of the Alloy CDA 715 was also resistant to SCC in all other environ.

NRC 1his Branch Technical Position clanfies the methods to rnents evetuated including NANO (2) at concentrations up to 1M, meet the requirements of 10 CFR Part 61 Samples of LOMI de-Alloy CDA 102 underwent SCC in NANO (2) environments at contamination waste stream resins solidified in Portland cement-concentrations as low as 200 ppm.

and unsohdtfied waste stream resins were obtained from the Peach Bottom commerciel nuclear power station. The radioiso-NUREG/CR 5720: MOTOR OPERATED VALVE RESEARCH tcpic compositron of the waste stream xas determined, and the UPDATE. STEELE,R; WATKINS J C.; DEWALL,K.G; et al sohdited samples were leached in deminerakzed and simulated EG&G Idaho, Inc. June 1992 103pp. 9207140288 EGG-2643 sea water. The compressive strengths of sarnples immersed for 62371:129 90 days in the two teachants was determined in addition to the This report provides an update on the valve research spon-samples otitained from Peach Bottom, filter sludge waste sored by the U S. Nuclear Regulatory Commission (NRC) that is stream and sohdeed waste forms were obtained from Nine M le being conducted at the Idaho Nationa: Engineenng Laboratory Point nuctear power station. The radiomotopic composation of The update focuses on the information applicable to the follow-the Nine Mde Point waste stream was deterrnined ing requests from the NRC staff. (1) examino the use of in situ

~ _. - _ _ _. _ _ _ _ _ _. - _ -

i 8

Main Citations and At>stracts test resu ts to estimate the response of a vane al dos'gneasis NUREQ/CR 5745; ASSESSMENT OF ISLOCA RISK METHODOL.

r a

conditions, (2) examine the rnethods used by industry to predict OGy AND APPLICATION TO A COMDUSTION ENGINEERING required valve stem forces of torques, (3) identify guidelines for PLANT, KELLY,0.La AUFLICKJL; HANEV,LN EG&G Idaho, satisfactory performance of motor operated valve diagnosbc5 lac. April 1992. 367pp 9205180177. LGG-2650. 61693 085..

systems, and (4) part;cipate en wnting a perfortnance standard inter system ioss-of coclant accidents (ISLOCAs) have been or gudance document for at;ceptable design bases tests The identified as important contributors to offsite risk for some nu---

authnrs have reviewed past, currunt, and ongoing research pro-clear power pl ants. A methodology has been developed for grams to provide the information avastable to address these thentifying and evalu.. ting plant-spectfsc hardware designs, items human factors issues, and accident consequence factors role-vant to the estimation of ISLOCA core damage frequency and NURE0/CH 6731: F;EDMONT SEISMIC RERECTION STUDY; A nak. This report presents a detailed descr ption of the applica-PROGdAM INTEGRATED WMH TECTONICS TO PRODE THE tion of this analysis fnethodology to a Combustion Engineenng CAUSE OF EASTERN SEISMICITY; GLOVER.L.; CORUH.C4 plant.

COSTAIN.J Ka et at. Virginia Polytechnic institute & State Univ.,

Blachburg VA March 1992;1a6pp U204300129 61500.001.

NUREO/CR 6785: EXPERIMENTAL RESULTS OF TESTS TO IN-A new tectonic model of the Appalachian orogen indicates VESilGATE FLAW DEHAVIOR OF MECHANICALLY LOADED that only one terrane boundary is present n the Piedmont and STAINLESS STEEL CLAD PLATES. ISKANDER.SK; Diue Ridge of the central and southern Appalachians. This ter.

RODINSON.G.Ca CORWIN W R.; et al. Oak Ridge National Lab.

rane boundary is the Taconic suture, which has been transport-oratory, Aptli 1992. 237pp. 9204300072. ORNL TM-11950.

ed in the allochthonous Blue Ridge / Piedmont crystalline thrust 61501:001.

nappe and is repeated at the turface by faulting arKI folding A strail crack near the inner surface of clad nuclear reactor caused by later Paleozoic orogenics. The suture passes through pressure vessels is an important consideration in the safety as-the lower crust and lithosphere somewhere east of Richmond 11 sessment of the structural integnty of the vessel. Four point is spatially asr.ociated with seismicity in the central Virginia seis-bend tests on large plate specimey.W clad and two unclad, mic zone, but is not comformable with earthquake focal planes were performed to determ% me eMeci nf stainless steel clad.

and appears to have httle causal relation to their localashon.

ding upon the permation of small sunace cracks aublected to Subsurface attucture in the central Virginta seismic n,,ne differs stress states similar to those produced by pressurlied thermal from that along sinke in the asoismic Roanoke River traverse.

shock conditions. Results of tests at temperaturen,10 and 60 The metamorphic plate is 9 km thick m the untral Virginia seis-degrees C belc'w the nikductshty transition tempe,ature have mec rone but only 3 km tNck near.he Roanoke Arvor, The cen-shown that a tough surface lay it Nsmposad of cladding and tral Virginia seismic tone may be more wrvasively broken by heat affected zone has arresied running flaws in clad piates distnbuted high angle ",ormal faults thin the Roanoke River under cu lions where unclad plates have ruptured. Further-area. Preliminary atterr. s to fit a single regional pants to all of more. the load-beanog o$pacity of clad platet with large sub-the planes of Munsey 5 3d Bollinger (. 85) give an appa'ently clad Jaws signtficantly exceeded that of en enciad plate with a -

good fit for a N55 degrees E trending p-aus. This is subparallel much smaller flaw. More testir g is necessary to unambiguously to the dominant NE regional p axis west of the Appalachians single out whether it is the cladding or the heat-affweted 20n0 that is pomarily responsible for the observed enhanced load-NURE0/CR-5732:'ODINE CHEPOAL FORMS IN LWR SEVERE beanng capacity of plates The compressive stresses that limit-W".NTSAnal Report.

DEAHM,E.C.;

WrBER,C F.;

ed the depth to which the flaw could propagate are absent m a KR w1 A; et st. Oak Ridge National Laboratory Apnl 1992.

7 visurgation event. Nonetheless, the exps,iments snow that 5

Dipp 9205150007. ORNL/T M.11861. 61691:233-b surface layer is sufficientry tougn, it could prevent a flaw, Calculated data 1 rom seven severe accident sequences in near the surface, f om propagatng along the surface Tno flaw hght water reactor plants were used to assess the chemical could tunnel below the surface, but a sufficiently tough sudace forms of iodine in containment. In most of the calculations for layer would reduce the maxirrum stress intensity factor, the seven sequences,60 dine entering containmerit from the re-actor coolant system was almost entirely m the torrr of CY with NUREG/081579h IMPACTS BRCNERSION 2.1 Codes And Data -

very small contnbutions of I or Hl. The largest traction of iodine Venficatiort RAO,R R.; KOZAK,M.W. Sandia Nabonal Laborato-in forms other than Csl was a total of 3 25, as l plus Hl. Within nes. ROLLSTIN.J.A.; et alc GRAM, Inc. Aprd 199L 206pp, the containment, 'he Cal will deposit onto walls and other surs 9204230141, SAND 912226. 61436 001, faces, as well as in water pools, largely tr) the form of lodide (t(.

IMPACTS BRC is a genere radiological assessment code that

)). The rada'ionanduced convers:on of IH in water pools into allows calcutation of potentia' impacts to nalmum individuals,-

j f(2) is strorg,y dependent on pH In sysd ms whete the pH was waste disposal wo4 ors, and the general populat90 resutting controlled above 7, little addalonal elemental iodine would be koni exemotion of a ry low level fr:dioactive wastes from regu-

{

produced in the containment atmomhere. When the pH falls fatory cornrol. The. coda allows calculatone to be made of j

below 7, however, it n,ay be assumod that it is not Long con-human exposure to the waste by many pathways and exposure trolled and large fractions of icdiin pa P 4tnin the contas scenanos. This document descobes the code history and the ment atmosphore may be produced quality assurance work that has been camed out on IMPACTS-BRC. The repm1 inch. des a summary of all the bteratura reviews NUREG/CR 5744: ASSESSMENT OF ISu A RISKm 00L-pertaining to MPAL SBRU up to Vorsion 2.0, The new code OGY AND APPLICATION TO A WEP NGHOUSE FGH-LOOP end data venficauon wrk necessey to produce IMPACTS BRC,-

ICE - CONDENSEH PLANT. ALY,DL; AtlFLlCKJ L:

Vers!on 23 is presented. General comments about the mortels HANEY.L N EG&G IdaN, Inc. Aprd 1992, 490pp 1?205180190.

and treatment r,f uncertainty in IMPACTS-BRC are also even.

EGG.2649. 61f 94:092.

Inter-System loss-of coolant accidents (ISLOCAs) have been NUREG/CR 5805: IDENTIFICATION AND ASSESSMENT OF identified as important contributors to offsito risk for some nu-CONTAINMENT AND RELEASE MANAGEMENT STRATEGIES c) ear power plants _- A methodology has boon developed for FOR A isWR MARK 11 CCNTAINMENT, UN.CG LEHNER J)l identifying and evaluating plant specshc - hardware designs.

Brookhaven Pational Laboratory. June 1 1002.

126pp.

human factors issues, and accident conseqvrwe factors role-9207060061. BNbh 7EG 52306. 62195:124..

vant m the est;mahon of iSLOCA core damage "eNency and '

Accident managartient strategon that have the potential to tri This report presents a detailed deset.ption of the appbcec maintain containment integrity 'and control or mibgate the re.

hon of this analysis methodology to a Westingharse four loop lease of radaoactivity fcilowing a severe eccident at a bojkng.

ice condenser plant.

water reactor with a Mack Il type of contaviment are identified

", -,,,, - -, - -. a -. _ - - -

~. -..

= - ---

Main Citations and Abstracts 9

and evaluated The stratepes are ' Mermd to as containment cirdes in so* age sludge dunng ets treatment and disposat The and release strategies Using infemanon available from prob-mapnty of the deterministic results from this evaluation in$ cat-ab*stic nsk assessments and other ecstir.g severe accdont re-ed a comfortable margin between the prudently conservatrve s arch, aM emplogng sitriphf.ed containment and release trees.

estimates of afinual doses and apphcabit pomussit10 levels.

Nin _ report identit.e6 the challenges a Mark 11 containment may Using Laf') Hypercube samphng methods, a stochastic uncer.

encounter dunny a severe accident, the mechanisms behind tainty and sensitsvity analysis was c inducted to estabhsh poten-those challenges, and the st'ategies that could be used to m>te tial ranges over which individual doses may vary and to identdy gate the challenges Py rneans of g safety othective tree, the the most sensetr o parameters and assumptions used in the v

strategies are linsed to the general safety obpctwos of contain-analysen Several exposure situations an'J radionucides have ment and release management As part of the assessment been #dentifed in this report for which the potential doses were process. the strateget are apphed to certain severe accdent calculated to be greates than an individual dotse level of 10 sequence categorios deemod important to a Ma4 Il contain.

mrem /yr if disposal of westes #n sewer systems approached 1 j

ment These sequence categones exhibit one or rnore of the Ci/yr, The results of this generic tt #dy will be ULed by NAC l

following characteristics hsgh protmbihty of core damage. high staff in doecting future modeling efforts supporting their final l

consequences, lead in a number of challenges, and involve the policy.

faJure of muittple systems The L6menck Generating Station is med as a representative Mark 11 plant to illut trate plant specil-NUREG/CR 5820: CONSEQUENCES OF THE LOSS OF THE l

ncs in this report RESIDUAt HEAT REMOVAL SYSTEMS IN PRESSURIZED

]

NUREG/CF4 5008; APPLICATION OF CONTAINMEN1 AND RE, WATER REACTORS. WARDLWa ARCIERI.W Ga HEATHS.

j LEASE MANAGEMENT STRAlEGIES TO PWR DRY-CON-EGM Idaho. Inc. May W2.183pp. 92062N N2M

.1 TAINMENT PLANTS YANG.J W; LEHNER.J R Brookhaven 021211fs0-National Lahorntory. June 1992 99pp 9207060007. BNL.

Dunng shutdown at Vogtle Unit 1 on March 20,1990, the loss NUREG42307. 62tM 001.

rA vital ac power and the Residual Heat Removal System This report identibes and evaluates accident rnanagement (RHRS) focused much attention on the need to evaluate system atrateg es that ars potentially of value in maintaining ctentain.

performance following such an event in hght water teactor ment miegnty and controlhng the release of radioactrvity fobow, (LWR) facihties The RELAP5/ MOD 3 traninent rion equilibnum Ing a severe accident at a pressunted water teactor with large.

System performance code ano an alternate methodology were dry r;ontainment The strategies are identified unirg a logic tree used to evaluate this scenano and to invepgate the accident

$1ructure leadong from the sately objectives and safety func.

moquences and identify key phenomenological and system j

hons, through the mochanisms that challenge thet6e safety func, behaviors charactermng these events. To investigate thormat tion ( to the strc >gies The stratagios are apphed to severe ac.

hydraulic behavior following a loss of the RHRS, studies evalu-i cdent sequences which have one or more of the following char.

ated the use of the steam generators as an alternatrve for re-actoristcs. vgndicant prot abikty of core damage, high conso.

movint' decay heat from mid-loop operatiort Additional studies quonces, gwe nse to a nurnver of potential challenges, and in.

investigted the effects of decay power and changes in reactor clude the f ailure of importet safety systems, Zron and Surry are coolant system (RCS) wate' icvel on system behavior when at.

saiecttd as te represensve plants for the atmosphenc and tempting to use the steam generators for heat removal. Undor trutwatmosoberm designs, respottrvoly.

these afternate heat removal conditions, analyses identified the ne to cwe uncovey in t% mnt a nom dam m N W i

MUREG/CR 580h IMPAOVEMENTS IN MOTOR OPERAT E D GATE VALVE DE SIGN AND PREDICTIDN MODELS FOR Ny rary thimble tube seat falls. Other evaluatons included assess-ing the impact of a i ss of the RHRS with the reactor vessel CLEAR POWER PLANT SYS1 EMS SBIR PhMe i Final dernals in place ond the uppor head removed, f or some plant j

Report. September 1090. Apnl 1991 WANG J K-; KALSI Kalsi En0ineenng. Inc, May 1992 116pp 920E260330 AEl'M S

"90' N0* # " # ""

"E# "#"#' **# #

hibit the' "downflow of water from the refuel pool cavity to the NO 1721 62121 001 lhis reserch en aimed at improving the pefformance of gate "O

ing d the fuM Lash, in N event boihng occurs in an o@n vaives at nuclear power plants (1) by developing improved RCS, the impact of the addition of borated water for extended models hr oporabihty prediction and (2) by identifying improve.

ments that overcome problems /kmitabons of the current de-signs Pharme t research is aime:1 at oeveloping improved operal-bonc acid twecipitation.

ng thrust models for the most common types of gate valves in NUREG/CR 5821: AUxtLIARY FEEDWATER SYSTEM RISK.

use at U S. nuclear power plants. Int.trumented valve test data BASED IMCTION GUIDE FOR THE KEWAUNEE NUCLEAR provided by Duke Power Company will be used to develop /

POWER PLANT, Lt OYD,R.C.; GORE,B F.; VO.T.V.; et al Bat-compare the analybcal predctions Specifcah, Phase ! '"-

talie Memorial institute, Pacdc Northwest Laboratory, Aprd -

i snawh will address shortco'rnngs in. the current techniques by 1992 29pp. 920514020L PNI, 7724,61668:128-investigating locahred contact stress 9s under disc tilting caused in a study sponsored by the U S. Nucieaf Re0utatory Com-l by fluid flow, by predicting inertial thrust overshoot, and by pro-mission (NRC), Pacihc Northwer,t Laboratory has developed and d

a comprehensive review of friction /galkng data for gate

~

ap led a methodology for denving plant-speCrfc Hsk based in-specton guidance for the aux #hary feedwater (AFW) system et fouREGNR 6814: EVAL U ATION OF EXPOSURE PATHWAYS TO pressunted water reactors that have not undergone penbabihtsc MAN FROM DISPOSAL OF RADiOACllVE MATERIALS INTO r sk assessment (PRA). This methodoogy uses * ~,, PH A re.

SANITARY SEWEn S\\ STEMS KENNE DY,W.Ea suits and plant operating esperience informate sting PRA.

PARKHURST,M A; AADERG R L; et al Beheile Memonal inste based inspection guidance information rect sloped for tute, Pacific Northwest Lk5 oratory. May 1932. 219pp.

the NRC for vanous mots was used to iden iric compo-92062$0328 PNL 7892. E,211'c146 neot failure modes This information was.

.bined with The dmenarge of radioactive matoriais to municipal sewer plant specific and industry wide comoownt in% ton and fail-systems is regulated by the U S. Nuclear Regulatory Commis-ure data to identify failure modes and fatture mechanisms for j

sion (NRC) in accordance wtth 10 CFR 20, or by agreement the AFW system at the selected plants. Kewaunee was select-states in accordance with additonal state regulations A generic ed as one of a series of piants for study. The product of this study using pathways. scenanos, dMa, and estumptions was effort is a pnonttred htting of AFW lailures which have occurred conducted by Pacife Northwest Laboratory (PNL) for the NRC.

at the plant and at other PWRs Tha hsting is intended for use to evaluate potential putAc dosos (fom esposure to redono.

by NRC inspectors in the preparabot. of 6ciptet.on plans ad-i

_ ~.,. _ _ _ _ -. _ _ _ _

__--__s

I i

i i

10 Main Citations and Abtf7 acts dressing AFW nskamportant components at the Kowaunce NUREG/C?l-5862: SCREENING METHODS FOR DEVELOPING l

plant-INTE RNAL PRES 3URE CAPACITIES F OR COMPONENTS IN SYSTEMS INTERF ACING WITH NUCLEAR POWER PLANT NUREO/CR4847: THE INF1UENCE OF PFdCOMPRESSION 04 1HE LOWER BOUND INITIAllON 100GHNESS OF A 533 0 REACTOR COOLANT 6YSTEMS WESLEY.D A EDE Enginner-eng Consultants (formerly EOF Engineering. Inc )

• EG&G RE ACT OR4GRAt;E STEEL.

DALLYJW, lHWIN.G R ;

idaho, fro. May 1992 103pp. 9206300237. EGGC673

< HANG,X4; et al Maryland. Unv of, College Park, MD. Mn 6MB7M' 1992,35pp 9206260337 ORNLSUD79 77784 62121:117.

Ruonhndam a Fuented for estabhbhing screerung 1hes report first desc9bes the test method emp4ying a pre' estimates of pressure capacities tor Itaid system correiorants compressed round t'ar subjected to em9act loading to mitiate a cleavage fracture The procedure to convert strain measure ~

subjected to interfacing System LOCA (ISLOCA) cte

, for t

nuclear power plant

• Includod in tto evaluation ase Jat ment into dyninic in$ation toughness K(Id) is described Also, enchangers, Idters. pumps, valves, flanged connet ? m urd i

the febutts of a fractograpNe analysis are correlated with the p po Tabular values are presented for stainless a,

.W featuras observed on the strain time traces, and techniques steel p;pe, as well as flanged connect ons with varic"s con ma-i used to disbngun.h initiation by en,ict cleavage or ductile tearing tonals and proloads. Simple analysis methods, togei,ser wib so I

are presented Micted i snabileties, are discussed for plant specific (,omponers NUREG/CR4865; TitCRMAL HYDRAUUC PROCESSES that must be evaluated independently. In addition to fatture or DURING REDUCED INVENTORY OPERATION WITH LOSS Of leak pressu.es, tabulated Icak rates or leak areas are includod RESIDUAL-HEAT RE MOVAL. NAF F,S A.,

JOHNSEN,G WJ where appheable PALMPOSE,D Ca et al EG&G Idaho, Inc A:

• 1992. 234pp.

9204300063. 0 GG-2671. 61501.2m NUREG/CR 5865: GENERIC SERVICE WATER SYSTEM RISK-Nuclear power plant conditions dunng outages differ markedly BASED INSPECllON GUIDE. STEWART,M A4 SMITH.C L from those prevaihng at normal full. power operabon on which EG&G idaho Inc. May 1992 21pp 0207060081, EGG-2674.

most past tesearch has concentiated This report 6 den %es the 6M W topics n(eid to onderstaf,;f pressunted water reactor re.

The risk-bat,ed enspection guide is intended to supplement sponse to an entended loss of, residual heat removal event U S. Nuclear Regulatory Comm6ss60ri (NRC) Temporary instrse.

dunny refuehng and maintenance outages. By identifying the tion 25t$/115. " Service Water System Operational perforrm posuble plant conditions and coohng methods that might be ance inspecton (SWSOPI)." The purpose of this guide is to ubed, the controlhng thermal-hydraulic processes and phenom, assist NRC enspection team loadors and team members to ena were identified. Gravity drain into the reactor coola it poontire inspection items and refine inspechon plans so their in-sybtem, core water boil +ff, and reflux coldensation cochng spections will address those elements that dominate the nsk as-wers investigated in detail for example plants from enh of the sociated with the service water system. This generic document three U S-pressuritej watet reactor vendors The reactor cool-prmnts nsk ensights obtained from probabikstic risk assess-ant systern pressure that would result from reflux coohng was monts and histoncal operating erponence. Because it as intend-calculated under vanous assumed conditions and compared to ed to assist tuspections at all commercial U.S. power teactors threshold pressures for vanous temporary closaes that might (which have wKlo vanations in service water system des 6gns),

te in use. The viability of va'ious potential gravity feed.and, some items may not be apphcable to every plant, Where posse-bleed approaches also was studed ble, the risk significance of the potential inspechon stems has been related to particular character.stics of plant design or enyt.

NUREQ/CR 5858;1NrORMATION FOR CONSIDERATION IN RE-ronmental conditions so that inspectors can determine which 4

VIEWING GROUNDWATEP "ROTECTION PLANS FOR URA-items may be applicable to a specific plant.

NIUM MU. TAILING 3 SITEb THORNE.P D Battelle Memorial lostttute. Pacihe Northwest Laboratory. May 1992 $0pp NUREG/CR 5866: FAULT TREE. EVENT TREE AND PIPING & IN-92060301E PNL 7986. 6tBp4148 STRUMENT ATION DIAGRAM (FEP) - EDITORSNERSION Quidehn, and acceptarse entona were developed for re.

4.0, Reference Manual.

MCKAY,M K.;

SKINNER fi L:

viewing cet ein aspects of ground water protection plans for WOOD.S T. EG&G Idaho, Inc. May 1992.144pp. 0206250352.

uranium mia taihngs sites The aspects covered include-1)

EGGt625. 62112:005.

leaching anc long-teim releases of hazardous and radioactive The Fautt Tree, Event. Tree, and Piping & Instrumentation Dia-constituents trom talhngs and other contaminated matenals,2J gram (FEP) editors allow the user to graphically buad and edit attenuation of huzardous and radioactive constituents in growed fault trees, event troos, and piping & instrumentation daagrams water under saturated and unsaturated conditions. 3) design (PatDs) The software is designed to enable the use of graphi.

and trtplomontation of ground water monitoring proyama, 4) cal-based editors found in the integrated Reliability and Risk As.

design and construction of ground water protection barners, and sessment System (IRRAS) FEP is made up of three separete

5) efficioney and effectfveness of ground water cicanup pro.

editors (Fautt Tree, Event Tree, and Piping & Instrumentation grams. The objechvo of these godelires is to assist the U S.

Diagram) and a utthty module. This reference manud provides a Nuclear Regulatory Commission staff in reviewing remedial screen-by screen walkthrough of the entire FEP System ;

achon plans for inactive waste sites and heensing appkcation documents for active commercial uranium and thonum milts.

NUREG/CR-5870: RESULTS OF LWR-SNUBBER AGING RE, SEARCH, BROWN.Of t L 6 i Engineenng Co. WERRY,EN.;

NUREG/CR 5860: : FRACTURE-MECHANICS DASED FArLURE BLAHNIK,0 E. Battelle Memorial Institute, Pacific Northwest ANALYSIS. ROSENFIELD,A Ra MARSCHALL,C.W. Battelle Me-Laboratory. May 1992. 199prt 9206120267, PNL 8051.

motsat institute.

• Oak Ridge Nationat Laboratory. June 1992 62038 001.

30pp 9207140031. ORNLSUBB2176241. 62373 227.

This report desenbes the aging Josearch results and recom-Twenty case studies involving the application of fracture me-mandations for snubbers used in commercial nuclear power chanics to structural antegrity have beet' reviewed and com-plants. Snubbots are safety-related devices used to restrain un-pared with a similar report pubbshed in 1970. Sixteen of the desirable dynamic loads at vanous piping and equipment loca-new cases discuss failures, while four are fitness-for purpose lions in t uclear power plants (NPPs). Each snubber must ac-ana'yses (ie,, evaluation of safe operating conditions of defect-commodate a plant's normal thermal movements and be capa-l containing structurest Co% tred with the earher study, no riig-ble of restraining the maximum off.norrnal dynamic toads, such nihcant improvement in accuracy of failure analysis was detect-as a seismic event or transient postulated for its specif c loca-ed However, expert opinion suggests that there has been s4g, tion. The effects of snubber aging and the factors that contrib-tyhcant improvement in. fatness.for purpose anaiysis-ute to the degradation of their safety performance need to be

. _. - - - _ - - - -. - - -. - _. _ - -. _ ~ - - ---

l

\\

l t

1 4

1 Main Citations and Abstracts 11 i

better undurstriod Thus Phase 11 Of Nuclear Plant Aging Re.

the product.on and use of RADS are reviewed and evaluated, search was condAtod to enhance the understanding of snuto and recommendations are made on dosage cahbrations, expo.

bot aq+ng and its coru,equences Pacific Northwest Laboratory sure control, monitonng, and personnel requirements. Special r

staff and thost subcontractors, take Enginoenng and Wyle Lab-emphasis is placed on dose cahbrators because these instru.

[

of atones, visited eight sites (encompassing thirteen plants) to rtients are used entensively to encature the dosages of radio.

r corviact interviews with NPP staff and to collect data on snub.

phartnaCouticals to be administered to pationts lhe difficulties but aging, testing, and rnanntenance, The Phase 11 feboarch of using dose cakbrators to quantify dosages of beta < and rnothodology, evaluation, results, concluwons, and recommere alpha. emitters are discussed. The advantages and disadvan-dations are desenbod in the report. Effective methods for serv-tages of using other instrufnents are es.amined, ar.d fecommers ico hfe rnonitor6ng of snubbof t are includod 6n the recommenda-dations are made on the types of instruments to be used for d6 tions ferent applications.

f NUREG/CR4871 V01: DEVELOPMENT OF EOU!" MENT PA.

NUREG/CR488k METALLURGICAL EVALUATION OF WELD RAMETER TOLERANCES FOR THE ULTRASONIC INSPEC-OVERLAID PlPE SECTIONS FROM BRUNSWICK UNIT 2 NU-TION Of STEEL COMPONENTS Apphcation To Componont" CLEAR POWER STATION. CZAJKOWSKI C.J ; BOWERMAN Ba Up To 3 inchos Thick GRE EN.E.R ; DOCT OR.S R.;

SCHUSTER.M191 al Brookha en National Laboratory. June 1

HOCKEY,4 La et al Dattelle Momonal Institute. Pacifo North

• 1992.124pp 9207140248. DNL fiUREG 52331. 02172.098, west Laboratory. June 1992, 07pp 9207140273 PNL4004-A metallurgical assessment of four sections t ' veld ovorlaid

62370232, This report documents work perfortnod at PNL on the effect pipe was portormed. The investigation consisted or strain Dage 2

measurornents, metallographic sectioning and mounting, scan-cd frequency domain equipment intoractons on the reliabihty of ning electron tnicroscopy, hardness and ferrite measurements, uttrasonic inservice inspectiott The pnmary focus of this work is radiography and dye penetrant examinations. A rov 6ew of the to provide information to the NRC on the acceptabihty of equip' fabocation history and onginal preservice and inservice exami-ment parameter tolorences as given in the ASME Doiler and nations was portormed and comparison was made to the actual.

Pressure Vessel Code Secton XI Appenda Vill Mathemstical cracks revealed attor sectioning in general, the report con-models woro developed for the entire ultrasonic inspection cludes that this wotd quality of the overlays was consistent with system including sound popagation through the inspection ASME quahty code class welds with adequate everage "as do-sample The models were used to determans worst cake inspec*

posited" forrite readings of FN>T, The chemical analysis of the tion scenanos for thin sections (piping), and those worst case welds were riormal for the alloys uWd (type 304 stainless steel, 6nspectron scenarios were then usnd in sensitivity studies to de-type ;r08 weld metal) Tho study also concludes that the ultra-termino the suitabihty of equipment parameter tolerances. Ultra' sonic inspection techniques used for inservice inspecton of the sonics htorature was reviewod to find worst care inspnction sce-overlays mar Not accurately depict the top 25% of the pipe in nanos outsido lho scope of Ib wel used, but none that were all cases and that crack growth is possible after weld overlay significantly worse wero found Expenments were pe formod to under certain conditons.

Confirm the important modeling results. The model predicted that ASME Code tolerances for equipment bandedth are ac.

NUREG/CR 5892: A HIGHWAY ACCIDENT INVOLVING UNIRRA-coptable, but tolerances for conto fregeency are too broad to DIAT ED NUCLEAR FUEL IN provido rehable inspecten of worst ~ase defects umng narrow SPRINGFIELD. MASSACHUSETTS,0N DECEMBEA 16,1991.

band systems. Esponments confirmod the basic trends predict-CARLSON,R.W,; FISCHERLE. Law *ence Livermore Natonal ed by the model, but the model shows greater sensit nty than is Laboratory. June 1992.106pp. 9201140243. UCAL ID 110638 found empincally.

02397:001 NUREG/CR4875: EXPERIMENTAL MODELING Of HEAT AND in the eqrly morning of Dec 16,1991, a sovere accedont oc.

MASS TRANSFER IN A TWO-FLUID DUBBLING POOL WITH curred when a passenger vehicle travehng in the wrong diroc.

APPLICATION TO MOLTEN CORE. CONCRETE INTERAC, ton colkded with a tractor traitor carrying 24 unitradiated nuclear TiONS GREEME,G A Drookhaven Natonal Laboratory. June fuoi assornbhes in 12 containcts on interstate 141 in Spnngfield, 1992,107pp 9?07140266 DNbNUREG 52325 62371:299.

Massachusetts The purpose of this report is to document the This report descrtbos the rotafts of seven sones of experb mechanical circumstances of the bevere accident, confirm tNe ments conducted to mvestigate heat and mass transfer phe, nature and quantity of the radioactive Vterials involved, and nomena m multicomponont bubbkng pools with apphcation to asuss the physical environmont to - ;h the containers were the rnodeling of interlayer heat and enass transfor between im-exposed and the response of the con nors and their ccntents miscible hqued layers and interfacial heat transfer to vertical and The report consists of five major sections. The first section do-honzontal boundanes for the CORCON computer code. Cntona scribos the circumstances and conditions of the accident and int the onset of entrainment between immiscible hquids, as well the finding of facts The second desenbes the containers, the as the rate of entrainment and the rate of setthng are devel, unitradiated nuclear fuel Assembhos, and the tie down arrange-Oped, which are apphcable for modchng of mass transport ment used for the trailor. The third desenbes the damage sus-dunng core concrete interactons. Host transfer models are de.

tained during the accident to the tractor, trailor, containers, and veloped for the case of stratifed layers as well as the case with the brurrad.ated nuclear fuel assembhos The fourth evaluates mase critrainmont between the layers. Finally, models for heat the accatont environment and its effects on the containers and tr&asfer with bubbkng to hontontal, dolled surfaces as well as their contents The final section g+ves conclusions donved from bubbling along vertical surtacos are presented which are appro, the analysis and tho fact finding investigaton. Dunng this severe pnate for boundary heat transfer analyses danng molten core.

accident, only minor injurios occurred, and at no 14mo was the concrete intaractom pubhc heatth and bar ty at risk-e NUREG/CR 5870 ASPECTS OF MONITORING AND OUAlfTY NU5tEG/84 0027; TRAC-PFt/ MOD 1 CALCULATIONS OF LOFT ASSURANCE FOR-RADIOLABELED ANTIBODtES EXPMMENT LP.02 6.

COCDINGTON,P.; GILL.C. Winfnth DARBER,D E. Mini eseta. Unrv of Minneapohs. MN

• Drookha-Technology Centro (Unsted Ktngdom) April 1992 280pp.

von National Laboratory-June 199147pp. 9207140262 BNL-9204230151. AEEWWl2464 61436 207.

NUREG 52328 62372 046.

Expenment LP-02-6 was the hrst large braak (200% Douole-This repet is Inten@d as an informatonal resource and Erded Cold teg) Loss-of Coolant Accident (LOCA expenment i

l guide for the U S. Nuclear Regulatory Commisson (NRC) and camed out in the Loss Of Fluid Test (LOF1) facihty under the NRC heensees who produco 'or use radiolobeled antibodies auspeces of the Organitation For Econom+c Cooperaton and (RADS) Components of quahty assurance programs related to Development (OECD) This report contains a companson of the

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U Main Citations and Abstracts results of egenment LP 024 with those of TRAC.Pr i/ MODI NURE G/l A-0045: ASSESSMENT OF REL AP5/MOE2 USING calculation LOCE i ARGE DREAK LO55-OLCOOL ANT EXPERNE NT L2-b b;

NUREG/lA 003(c ANAlvstS Of LODI TEST DLO2 OHREE PER-0' "O

CENT COLD LEG BREAK) WITH RELAPS CODE LCRIVE N. A H Cent'al Electnuty Generating Board March ort documents the resuits and conclusions of the 1992. 7Dpp 9204300111. U RD/L/3iO9/Rbt f.,1500157.

RELAPS/ MOD 2 code assessment in the analysis of LOCE Test The tet.t BLO2 was a United hngdom specified test conduct-

,, 5 1 v m messment study is te Movide sys-d 2 loop test system at Ispra. North tomatic assessment of RELAPS/ MOD 2 Code reli..ve to code ed on the elecincatty heatrlB' senes of tests sponsored by the

.taly, as part of the LOui development code improvement and the enhancement of user Jmnt European Researm Commistoon The protest calcuistions 9#d"h"C' for this test were portormed with the RELAPS/ MOD 1 codc and the post tat anaiyos was named out uung P6 LAPL/ MOD 2 NUREG/lA-004G: ASSESSMENT OF RELAP5/ MOD 2 USING SE-Compensons between the code predictons ano the test data MISCALE LARGE BRE AK LOSS OF COOLANT EXPERIMENT are given. and for the case of the post test MOD 2 calculation, S 0fr3 LIANGES ; KAOL; CHIOU.J-L ; et al. Institute of Nu-detailt 1 stu3es of to rodes portormance in a number of veas c ie,,

E nergy Research (Tainan) Apnl 1992 176pp are incluM D206050049 61891.166

-~

NURE G/lA-0037: ASSESSMENT OF RE LAPS / MOD 2. CYCLE This report presents the results of the REL APS/ MOD 2 post-36 04 AGAINST LOFT SMAll DREAK DrERIMENT L3 5 test assessment utikting a semiscale large break loss-of coolard E RIKSSON.J Swxhh Nocicat Dower inspectorate (Statens capenment numbered 5 06-3 Emphasis was placed on the ca-Karnbeftinspektion) March 1992 97pr 9204300097 STUDS.

pabihty of the code to calculate break flow rates during system VIK/NP843 6151*> 245 blowdosn stage, emergency core cooling system (ECCS) injec-An independent aswssment of ( o k ? LAPS / MOD 2 code was tion bvpass dunng refitt stage. quenching dunng reflood stage.

conducted by Studsvik Energetekna AB The LOF T small breah and peak clad 4ng temperature behavior throughov! the whole openment L3 5 was amssed uwng the RELAPS/ MOD 2 code.

capenment a

Three calculations were carned out one base case calculation and two s.enotivity calculations with modal changes The tran.

NUREG/lA-0049: THERMAL HYDRAULIC POST. TEST ANALY.

SIS OF OECD LOFT LPJP 2 EXPER! MENT. PEN A JJ; Sent piedictions compare reasonably well with the rapunment as regards firsthand parameters such as system pressures and ENCISO,5 ; REVENTOS F, Conscio de Segundad Nuclear fluid temperatures Vanatsons are enumerated s.nd discussed (Spain) Apnl 199t. 207pp 920515009t ICSP-L P-F P-2.

61692 160.

NUREG/lA-0036: ASSESSMENT OF 1RAC-PF 1/ MOD 1 AGAINST An assessment of RELAP5/ MOD 2 and SCDAP/ MODI AN INADVLRTLN1 F EEDWATER L'NE ISOLATION TRAN-against the OECD LOFT expenment LP.FP.2 is presented LP-SIENT IN THE RINGHALS 4 POWER PLANT. SJODERG.A FP--2 studies the hypothetical release of fission prod 1 cts and Swedish Nuclear Power inspectorate (Statens Kamkraftinspek-their transport following a large tweak LOCA scenano. The 1.M). March IOM ts2pp. 9204300090 S T UDSVI ANP08'01 -

report composes a pneral desenption of the LP FP-2 empen-6iT21R ment, a summary of therinal. hydraulic data, a simulation of the An inadvertent feedwater line isolation tranuent in a inice L P.FP.2 expenment, results of the RELAP5/ MOD 2 base calcu-loop Westinghouse PWR has been Omulated with th(i frozen lat:0n, the RELAPS/ MOD 2 senotivity analysis, the SC,DAP/

version of the TRAC-PF1/ MODI computer code The results MOD 1 nodabration for an LP. FP-2 expenment, the results of tenated the capacity of the code to quantitatively predict the the SCDAP/ MODI calculation, and the summary and conclu.

difieroni pertinent phenomena for accurate predictions of the g,ong system response, particular care was reauned in the nodaliza-tion of the steam genciator dawncomer and restnctions had in NUREG/lA 0050: TRAC-PF1 CODE ASSESSMENT USING OECD t e imposed on the allowable manimum tme step-LOFT LPJ P 1 EXPERIMENT. BARBERO F.J Consejo de Se-NUREG/lA-0043; ASSESSMENT STUDY OF RELAPL/ MOD 2 gundad Nuc. ear (Spain). Apol 1992.185pp 9206050036. ICSP-LP FP 1. 61891:001.

CYCLE 36 04 BASED ON THE DOEL 4 MANUAL LOSS OF This report assesses thermal hydraube aspects of LOFT LP.

LOAD TEST OF NOVEMBER 23.

1985 STUDDE.E J ;

FP 1 expenment making usa of TRAC-Pfi/MODt. LP-FP.1 on.

DESCHU T1 E R.P.

TRACTEBEL Mo ch 1992 33pp r

penment studees the system turmal-hydraubc and core thermal D204300f 06 61500 229

'05P0050 ID' i"'tial ana boundary cond tions similar to a large-The loss of enternal load test conducted on the DOEL 4 break design basm '.DCA leadsng to fission product rt ; ease power plant has been analyzed on the bas 2s of a high quakt, kom the Not cladhg gap region. It also assesses the fisuon data acqumt on system A detailed numencat anaNos of the product retention effectiveness of the PWR LCCS 40 best esti-transient by means of the best estimato code RELAb/ MOD 2 is mate Conditions.

pretented The RELAP5 code is capable of neulating the baMc plant t+ haviour. Deficiencies noted involved structural heat sim' NUREG/lA-0057: ASSESSMENT STUDY OF REL%P5/ MOD 2 ulation, acoustic phenomena, and excessive interphase drag CYCLE 36 04 BASED ON THE COMMISSIONING TEST REAC-NUREG/lA 0044: ASSESSMENT STUDY OF RELAD5/ MOD 2 TOR TRIP AT FULL LOAD AT THE PHILLIPSBURG 2 NUCLE-CYCLE 36 05 0ASED ON THE TlHANGE 2 REACTOR TRIP OF AR POWER PLANT. GERTH G. Siemens AG Bereich Ener.

JANUARY 11.1983 ROUELG P ; STUDDE.E J TRACTEDEL gieerzeugung (KWU) Apot 1992. 83pp.

9205180172 March 1992 103pp. 9204230203. 61435 001 61740 014.

As part of a commissioning test senes. a plant inp from The commissioning test " Reactor Tno at Full Load". which 100". Power was performed at the Tihange-2 Nuclear Power was performed at the nuclear power plant Phthppsburg 2 (KKP.

Piant A simulation of this transient was pertormed by means of

2) was recalculated with RELAPS/ MOO 2, The companson of the code RELAPS/ MOD 2/ Cycle 36 05 The assessment of the the results wes became sufficiently uncovered Report it was ong(nally des 6gned for loss of-Coolant Accidents, foe pnmary to secW lary heat transfer to be signibcantly re.

but is now finding wider apphcations. RELAP5/ MODI was used duced. The ensuinh.wiat tsp of the pomary fluid resulted in a lot a pro test calculation, and RELAP5/ MOD 2 for the detailed reduction m power induced by the modorator feedback. The pn, post test analysis. This was to allow cross-code comparisons, (nary system pressure increased to the safety relief valve set.

assess the possibility of using RELAP5 for pressurized tran-potrit tiefore the fall in reactor power allowed the raismatch be.

sients and because the final phase of blood and feed which twoon pnmary 1,. tem heat input and heat removal via the occurs in test S102 is rnora representative of Small Break tran-steam ger.erator to be accommodated by cycling of the pilot op-s'ents than F,essunzed f aults, This report documents the re-etated reliot valve ("OHV) Companson betwoon calculation suits of this calculaton and comparisons with the test data.

and data shows generally good agroornent, though with discrep.

After accounting for test cordtions and events outside the ongh ancies in some arent Weeknesses in thf. code's troutment of nel specification the RELAPS/ MOD 2 code was tvunti to perform interphase drag and in the representation of the prosaurizer rather weH spray are indicated, although a shortage of definitive data, par-NUREG/lA 0062: ASSESSMENT OF TRAC.pr1/MODt AGAINST ticularly in the steam Derierator, may also be a factor. The over, AN INADVERTENT PRESSURt2ER SPRAY TOTAL OPENING prediction of interphane drag lod to a tendency to underpredict TRANSIENT IN JOSE CABRERA POWER PLANT.

the in,tial inventory in the steam generator ano al* perhaps, to pvorpredict ce steam generator heat transfer while the tubos FANEGAS,R M. Technatom, S A. (Spain) April 1992. 70pp.

9206050019 ICSP-JC-SPR T. 0t BD2.214.

were being uncovered Thort is indication that the pressunter vapor region conditions were close 10 equilibrium dunng spray This teport assesses the use of the TRAC PFt/ MODI code to operation The point lonetics model th RELAP5/ MOD 2 proved a reproduce the actual transient that occurred at Jose Cabres -

viable means of representing the power history for this tran.

NuclJar Power Plant on August 30,1984 and is thus part of an sent assessment unorcise within the tramework of the ICAP program.

NUREG/lA 0059: ASSESSMENT OF RELAP5/ MOD 2 AGAINST NUREG/lA 0063: RELAP5/ MOD 2 CALCULATION Or Cmc 0

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NATURAL CIRCULATION EXPERIMENTS PERFORMED WITH LOFT TEST LP.FW 01. CROXFOD,M G.; HARWOOD.C.;

THE hEWET iti FACILITY, HYVARINEN.J.; KERVINEN,T. Tech-HALL,P.C. National Power Nuclear (Urnted Kingdom). April nicht Research Centre o, rinland (VTT). AprH t992. 73pp.

M2, epp. 92WM2m NEMR 6Kp2 9205140248 431660 226 RELAP5/ MOD 2 is teng used by GDCD for calculation of cor.

Natural circulation experiments camed out in the REWET ill tain sman break losoopant accidents and pressurtred tran-b#*#

facility in 1985 have been usod for RELAP5/ MOD 2 assess-ment The REWET til faellity is a sc 'ed-down model of VVER.

MOD 2 to model the pnmary food-and blood recovery procedure 440 type reactors The facility consists of a pressure vessel in following a complete loss-of feed ator event, post test calcula-tions have been carried out ol OECD LOFT test LP FW-01, This which the downcomor as simulated Wth an external pipe assem-bly, hot and cod legs with loop seals and a honzontal steam report descnbos the companson between the code calculations and the test data. It is found that although the standard version generator. The volume scating factor compared to the reference of RELAPS/ MOD 2 gwes a reasonable predicton of the experi-teactor is 1:2333 The present paper summanrea the expen-mental transient the long term pressure history is better calcu.

ences gained in the RELAPS/ MOD 2 calculations of selected lated with a modified code version containing a reyssed honton-i l

REWriT lH single and two-phase natural circulahnn expen' tal stratif caton ertainment model. The latter AHows an im-monts. The code's ability to represent the.Tiaan phenomena of prove alculation of entrainment of liquid from the hot leg into experiments in both caises r ws satisfactory.

the surge line. RELAPS/ MOD 2 is found to grve a more accurate NUMG/lA 0060: APPLICATION OF THE RELAPS/ MOD 2 CODE s UK s R TR N O /

TO THE LOFT TESTS L3 5 AND L3-6. SCRIVEN.A.H National Power (United Kingdom). AprH 1992. 51pp, 9206050025. ESTD/

NUREG/lA-0064: ANALYSIS OF SEMISCALE TEST S-LH 1 L/0117/R89 61892 286-USING RELAP5/ MOD 2 HALL.P.C.; BULL.D R Natonai Power

- Ar APS/ MOD 2 is being used by Nabonal Power Naclear, Nuclear (United Kingdom). April 1992, 40pp. 9206150263. GDI Technology Division for calculation of certain small break los.n-PE N/725. 62003:138.

of.cooiant accidents and prbasurized transients in the Sirewell The RELAP5/ MOD 2 code is teng used by GDCD for calcu-

'B' PWR The code version being used is RELAPS/ MOD 2 cycle lating Small Dreak Loss of Coolant Accidents (SBLOCA) and 36 05 Wintnth version E03 As part 31 the programme of so pressunred transant sequences for the Seewell

'B' PWR.

sessment of this code a number of compansons of calculations These calculations are boing camod out at the request of with entegral test facihty experiments are being carned out. At Sizewell 'O' Project Management Toam. To assist in validating the request of NPN.TD the LOFT 2.5*. small cold leg break RELAP5/ MOD 2 for the above application, the code is being tests L3 5 (pumps off) and L34 (pumps on) have been calculat-uwd by GDCD to model a number of smaH LOCA anu pressur-ed These p'evously performed tests' involved a number of fea-god fault simulaton experiments camed out in vanous integral tures, including stratif caton, pump performance and offtake of-test facehtees The present report desn*=a a RELAPS/ MOO 2 lects which suggestod they would be useful measures of code analysis of the sman LOCA %i SLH 1 which was performed performance.

on the Semiscale W N facility S. LSt simulated b small

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- 14 Main Citations and Abstracts LOCA taused by a bret in the co;d leg pipework of an area eling on major thermal hydraubc parameters, espcCally on peak equal to 5% of the cold leg flow area cladding temperature (PCT), These sensitivity items are. Single flow channel and single heat structure (Case A). two flow chate NUREG/lA-0066: ANAtYSIS OF SEMISCALL TEST S-LH 2 not and two heat structures (Case BL reflood option added USING RELAPS/ MOO ( DRODIE'P : HALL,P C. National Power Nuclear (Unrted hngdom)' April 19h2. 28pp' 9206150241 GD/

(Case C) and both reflood and gap conductance options added (C-e D) The code. RELAP5/ MOD 2 Cycle 28.04 with the base n adehng, predicted the key parameters of LOFT IBLOCA Test e EL

/

code is being used by National Fower L51 better than Oases A, D, C and D. Thus, tt is concluded that Nuclear Technology Dmsun for calculahng Small-B'eak Lobs, 00 Coolant Accidents (SDLOCA) and pressunted transient se, Die seg4 how channel indig W core is bem than Ve two flow channat mcdeling and two heat structure is also better Querices for the Sitewell V PWR. To assist en validating RELAPS/ MOD 2 for the above application, the code is being than a single heat structure modeling to predict PCT at the con-tral fuel tods. It is recommended to use the reflood option and j

used to rnodel a number of smatt LOCA and pfessurated fault simulation empenments carned out in integral test facihties.1he not to use gap conductance opton for this LS 1 type IDLOCA.

1 present report desenbr s a RELAP5/ MOD 2 analysis of the small NUREG/tA-0070 - ASSESSMENT OF RELAPS/ MOD 2 CYCLE LOCA test S-LH-2 which was performed on the Semistaie Mod-30 04 WITH LOFT LARGE BREAK LOCE L2-3. BANG,Y.S t 2C Facility. S LH.2 simulated a SBLOCA caused by a break in-KIM,H J Korea institute of Nuclear Safety. KIM.S H Korea the cold leg pipNork of an a ca equal to 5% of the cold leg Atomic Energy Research Institute. April 1992, 144pp.

f ow area RELAP5/ MOD 2 gave reasonably accurate predic' 9205159101 61092:044 tons of system thermal hydrauhe behawor but failed to calcu-The LOFT LOCE L2 3 was simulated using the RELAPS/

late the core dryout which occurred due to coolant bolboff pnor MOD 2 Cycle 36 04 code to assess its capabmty to predict the to accumu!ator ingcton The error is behoved due to combina-thermal-hydraube phenomena in LPLOCA of the PWR, The te.

tions of errors in calculating the liquid inventory in the core and actor vessel was bimulated with two core channe s and split downcomer modelling for a te case calculation using the grd n 1 in co e frozen code. From the results of the base case calculation, deft-NUREG/lA 0066: RELAPS/ MOD 2 ANALYSIS OF LOFT EXPERI.

ciencies of the critical flow modot and the CHF correlation at MENT L9 4. KEEVILLM R National Power Nuclear (United high flow rate ir ere identified, and the severeness of the rowet.

Kingdom). Apol 1992 38pp-9206150246. GD/PE N/721.

ting entena were aw found Additional calculation using an up-62063.051 dated version of RELAP5iMOD2 Cycle 36 04 including modith As part of a program to validate RELAP5/ MOD 2 for use in cations of the rewet cfrtena shows a substantial 6mprovement in the snaiysis of cenaan fault transienta in the Sirewell O PWR, the core thermal response, the code has been used to simulate emenment L9 4 comed out in the Loss Of. Fluid Test (LOFT) facility Enpenment L9 4 timu-NUREG/lA-0071: ANALYSIS OF THE UPTF SEPARATE EF-lated a Loss-Of Othat+ Power Anticipated Transient Without FECTS TEST 11 (STEAM-WATER COUNTERCURRENT FLOW Tnn (LOOP ATWT) in which power is lost to the pnniary coolant IN THE RROKEN LOOP ltOT LEG) USING RELAPS/ MOD 2-pumps and main feed as lost to the steam generators but the DiLLISTONE.M J. 'Mnfrith Technology Centre (United Kingdom).

control tods fall to insert in the reactor core, RELAPS/ MOD 2 June 1992. 31pp. 9207140017. AEEW M2555, 62399 001.

gocerally predicted the transient well, although there were some RELAPS/ MOD 2 predictions of countercurrent flow limitation in differences enmpared to the test data These differences are the UPTF Hot leg Separate Effects Test (test 11) are com-largely due to the use of power and flow as boundary conditions pared with the experimental data. The code under-estimates, by and because of ursortainties in tne power and flow expenmen-a factor of more than three, the gas flow necessary to prevent tal data The most noticeable difference wa4 that the steam hquid runback from the steam generator, and this is shown to generator was predicted to boil down too fast. Thst is believed be duo to an oversimphied flow te(me inap whlCh does not to be partty due to errors In the RELAPS interphase drag model-allow the possibility of stfatified tiow in the hot leg riser, The The RELAPS calculation also showed the pnmary pressure to predicted countercurrent flow is also shown to depend wrongly, be very sensitNe to the pnmary flow rete, rm kang the exact sim-on the dopth of liquid in the steam generator plenum. The same ulation of pnmary side rehef vaive movements difficult to repro-test to also modelled using a version of the code in which strati-l doce fied flow in tM riser is made possible. The gas flow needed to NUREG/lA-0068: ASSESSMENT OF THE "ONE FEEDWATER prevent hauid runback is then predicted quite well, but at ali PUMP TR69 TRANSIENT"IN COFRENTES NUCLEAR POWER lower gas flows the code predicts that the flow is completely PLANT WITH TRAC-DF1 CASTRILtO.F f4 droelectnca Espan.

unrestncted 6 e. Equid flows between tutt fiow and zero flow are o!a. NAVARRO.A G.; GALLEGO.L Union iberoamencana De not predicted. This is shown to happen because tha. code Tecnologia_ Apol 1992, 79pp 920f)D80039 ICSP-CO TURFW T.

cannot calculate correctly the hquid level in the hot leg, mainly because of a numencal effect of upwind dononng in the mo-61958 173 This report presents the results o me assessment of TRAC, mentum Hun terms of the code's basic equabons. It is also r

shown that the code cannot model the considerable effect of BP (GtJfl code with the mor.a the C, N Cofrentos for sim.

l

. ulatio of the transieni '..gnated by the manual top of one FW the ECCS injection pipe (which runs inside the hot leg) on the bound level l

pump.

NUREG/tA 0069: ASSESSMENT OF RELAP5/ MOD 2 CYCLE NUREG/lA-0072: LOFT INPUT DATASET REFERENCE DOCU-3604 USING LOFT INTERMEDIATE BREAK EXPERIMENT L5 MENT FOR RELAP6 VALIDATION STUDIES. BIRCHLEY,J C.

1. LEE.E.J : CHUNG.D D-; K!M,H.J. Korea Institute of Nuciear

' Wintnth Technology Centre (United Kingdom) April 1992. P5pp.

Saiety AprJ 1992.140pp 9205140233-. 61670.141.

9206050096. AEEW R2454,61894 001.

The LOFT intermediate break experiment L51, which stmu-Analyses of LOFT expenment data are' being carried out in tales 12 inch diameter ECC hno break in a typical PWR, has order to vahdate the RELAP5 computer code tv future apphca-been anaiyred using the reactor thermal /hydrauhc analysis tion to PWR p; ant analysis. The MOD 1 dstasef was atso used code RELAP5/ MOD 2, Cycle 36.04 The base calculation, which by CEGB Bamwood who subsequently converted the dataset to l

rnodeled the core with single flow channel and two heat str6>

ru, with MOD 2. The mod!hcations included changes to the to tures witht '

t. ing f

%ns of reflood and gap conductance MMt60n to talie advantage of the Crossflow junClion option al model hm ' te i s, - +

completed and compared with ed appropnate locations. Ad$tional pipework representation was penmenta Oata Senn..., studies were carried out to invesS introduced for breaks in the intact (or actrve) 1000. Further gate the effects of nodattration at reactor vessel and core -mod-changes -have-been made by Wintnth following discussion of i

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l Main Citations and Abstracts -

15

' calculations perfortned by the CEGO and Winfnth These con-rontal coeurrent stratifed steam / water flow in a rectangular Cern the ;legree of noding m the steam genefator, the fluid channel. A redahrstion senStivrty study and a $imulation with a volume of the steam genotator downcomer, and the location of fixed interfacial area, same as test section, were also camed the reactor vesset downcomer bypass path. This document 9-out to examine the effect of the interfacial heat and mass trens-scribes the datanet cor tents relating to the volierne, junction, for. The results showed that the RELAP5 cooo model under the end heat slab data for the 6ntact loop, reactor pressure vessel.

hontontal stratified flow regime predicted the condensation rate brMen loop, steam generator secondary, and ECC system Also well, though the interf acial heat transfer area was underpf edict-desertbed are the control system for steady state trutmitration, ed. However, some d*crepanc,es with expenmental results standard trip settings and boundary conditions.

wore found in wator layer thtckness and local beat transfer co-NUREG/lA-0073: TIME STEP AND MESH SIZE DEPENDENCIES edicient esscially when there was a wavy interf*^e. The inter-IN THE HEAT CONDUCTION SOL UTION OF A SEMI-facial wave structure was found to play an important role in fle-IMPLICliflNITE DIFFERENCE SCHEME FOR TRANSIENT scnb'ng the interfacial heat and mass 1ransfeu as obtained in TWO-PHASE FLOW. O MAHONEY,R. Winfnth Technology the erperiment.

Centre (United Kingdom) Aprit 1992 5 t pp. 9206050101.

NUREG/lA 0078: REVIEW AND

SUMMARY

OF TRAC ASSESS-AEEW M2590 61894.096' nd estabishes the causes of, previ-MENT FROM THE INTERNATIONAL CODE ASSESSMENT This report examines, a AND APPLICATION PROGRAM (ICAP) SCHNURH.N M. Los ously identified time step and mesh site dependencies These Alamos National Laboratory. Apot 1992. 257pp. 9206050091, dependencies were observed in the sc:uton of a coupled LA UR 91-2543. 01890.001.

system of heat conduction and fluid flow equatsons as used in TRAC PF1/ MOD 1 has been exercised by severalinternatior-the IRAC-PFt/ MOOT computor code The report shows that a a' uters as a part of the ICAP Program. Participants are re-

[

significant time step sire dependency can anse in calculations quested to prepare a report summanring the results of their of the quenttiing of a provinusly unwetted surface The cause work. These assessment repor1s contain discussions of the of this dependency is shown to tse the expbcat evaluation, and code accuracy, errors and deficiences, new usor guMelines, subsequent smoothing, of the term which couples the heat and recommendations for code upgrades and mod &ations.

transfer and fluid flow equations An asial mesh site dependen-cy to also vientified, but this is very much smpPer than the time NUREG/lA 0079: POST TEST ANAL) SIS OF LOD' TEST BT 12 step stre dependency. The report conclu*a tr.et the timo step USING RELAP5/ MOD 2. SMETHURST.AJ. Winfnth Technology size dependency rtgresents a potential (4 Station on the use of Centre (United Kingdom). Apnl 1992. 71pp. 920C050086c large time step gires for the types of cEdation discuesed This AEEW R2645. 61893 248.

limitation affects the prer.ent TRAC PF1/ MODI computer co(te This report desenbes calculations Jamed out with RELAPS/

and may similarly affect c'her semi-impi. cit finite difference MOD 2 on LOBt expenment BT-12. a large steam line txeak.

codes that employ similar techniques it is likely to be of great-The following sensitivity studies were performed heat losses on est significance in codes where mt' tiep tecnniques are used the intact steam generator; discharDe coefficient at break; waior to allow the use of large time steps.

in steam lines, nearly ;mplicit numencs. Quaktative'y the Deneral NUREG/lA-0074. RELAPS/MOU2 POST TEST CALCULATION trends of BT 12 were predicted well, in particular the timing of OF THE OECD LOFT EXPERIMENT LP SB t.

PERE 2,J ;

events was larly accurate.

MEND 12ABAL.R Consejo de Seguridad Nuclear (Spain) Apol NUREG/lA-0080: ASSESSMENT CC RELAPS/ MOD 2 USING SE-s 1992.86pp 9206050077. ICSP LP-SB+R 61893.002-MISCALE INTERMECIATE BREAK LOSS-OF COOLANT EX+

This document presents the analysis of the OECD LOFT LP.

PERIMENT S-IP4. CHOU,G H.; HORNG.T-S.: LIA,L Y. Institute SB 1 Exponment performed by the Consejo de Segundad Nu' of Nuclear Energy Research (Taiwan). June 1992, 210pp clear of Spain working group making use.of RELAP5/ MOD 2 in 9207140022. 62373.271.

f the frame of the Spanish LOFT Project LP SB-1 expenment This report presents the results of the RELAPS/ MOD 2 as-studios the effect of an early pump inp in a small break LOCA gesg, ut*2mg a Semiscale intermediate break loss-of cool-scenaao with a 3 inch equivalent diameter break in the hot leg ant expg mont S 10 3 Comprehensra anaysis with RELAP5/

of a commercial PWR MOD 2 ir performed to predict the tran,ient thermal-hydraulic re-NUREG/lA-0075: RELAPS/ MOD 2 ANALYSIS OF A POSTULAT.

sponses of the expenment. Test S fd 3 is a 21.7% communk ED " COLD LEG SBLOCA" S:MULTANEOUS TO A " TOTAL catrve cold leg break LOCA expenment using Semiscale Mod-BLACK OUT" EVENT IN THE JOSE CADRERA NUCLEAR 2A facMy in 1982, for the principal objective to provide refer.

STATION. REBOLLO.L Union Electnca renosa. S A April ence data for companson of Semiscale test results to LOBI fa-1992.24?pp 9206080045. lCFP-JC-SB3 R. 61962.07 cility B-RtM test results. Through extensive companaon be -

The RELAP5/ MOD 2/36.04 code has been used on a CYBER tween test data and best estimata EEL APS calculations, the ca-l 180/830 computer and the simulation includes the 6" RHRS pabilities of RELAPS/ MOD 2' to predict the immediate break charging line, the 2" pressunzer spray, and the 1.5" CVCS LOCA accident were assessed. Emphasis was located on the -

make up line pr 3 breaks. The assumpton of a " total black.

capabihty of the code to enlculate core level depression and

.sinci en with the occurrence of the event has break. flow rate dunng system blowdown, pump section liquid out condition" d t been rnade in order to consider a plant degraded conditon with

_ seals phenomena, and temperature excursions behavior etc.,.

total active failure of ECC& As a result of the analysis, esti-t%sghout the whole experiment. Besides some sensitivity stud-mates of the " time to Core overheating startup" as well as an ses involving the effect of steam generator secondary ssde pres-evaluation of alternate operator measures to mitigate the conse-sure boundary, adjustment of two phase discharge coefficient.

quences of the event have been obtained. Finally, a proposai

. Intact loop pornp coastdown behavior, and some interesting

~

for improving the LOCA emergency operating procedure (E-1) studies regaiding break flow etc. were also investigated in this has been suggested.

report.

NUREG/lA-0077:' RELAPS ASSESSMENT ON DIRECT CON.

NUREG/lA 0082: =THE ASSESSMENT OF RELAPS/ MOD 2 I

TACT CONDENSATION IN HORIZONTAL COCURRENT AGAINST IVO LOOP SEAL TESTS, KYMALAINEN,0. Imatran STRATIFIED FLOW, LEE.S.; KIM.H JL Korea instdute of Nuclear Voima Oy - (IVO) (Finland). Apol 1992. 36pp. 9206080024.

Safety; April 1992.110pp. 9205140240. 61696.066 61058 247, Assessment on the direct. contact condensation model was

. RELAP5/ MOD 2 analyses of a full scale and 1/10-scale at.

camed out using RELAP6/ MOD 2 Cycle 36 04 and the RELAPS/

mosphenc uer water loop seal facilbos have been Conducted.

MOD 3 Version Sm5 codes. The test data was obtained from the The calculations have been periormed with the version 36.05 expenr0ents at Northwestern University, which involved the hork and also with a modtfied version with the treatment of interfacial i

87 t

F*

5s 7V9'"*

W'F 'W k kf it 9

37fim g-t ph 1WTWw.'t-irW'*fT-

1. "-*g rtiess-se,--s3wemwy'vM+=n-?e mim-ter'.e-ee,+b.-uuwe'"

>'m u

au%---m-m.<*w-""-------------'ia---

--a L---=

16 Main Citations and Abstracts drag r. hanged in the loop seat beh The calculatoo residaal The sirnuletons of Marviken CFT 15 and 24 have beun per.

wate* level ditlers frorn that rraasured en the espenments, the formod usug RELAPS/ MOD 2. For the modeling of a norrte as cornputet.onal valen boing lower The gas superficial veloctty a pipe. the resuits of simulations and the CFT 15 test data are needed for loop stal cleanng is also predicted lower by in good agreement. but the Birmlations underpredict by about 5 RELAPS. The 6ntoriacial drag mooiticatone slightly improved the to 10% in transtten region between subcooled and two-phase results, but an agreement with the expenmental dat3 was found_

in the two-phase regon there hapfens the fluctuations of the calculated mass flowrate for the cabe of using the cntical flow NUREG/lA 0044: RELEVANT P8.SULTS OSTAINED IN THE model in RELAP5/ MOD 3. It seems that the Improvement of the ANALYSIS OF LOBl/ MOD 2 NATURAL CIRCULATION EXPERI-cnbeat flow model in RELAPS dunng the transttion pered is '

2 MENT A2J7A D'AURIA,F.; GALASSI,G M, Univeruta' Dogli necessary, RELAPS cnt: cal flow model underpredicts the CFT Studi Di Pisa. Pisa. Italy Aphi 1992.160pp 9205180160. NT 24 data by 10 to 20% in two phase choked flow region, wtute 163 (90) 51733.003 its predictions are in good agreement with subcooled woked The present document doscr6es the activities carned out at tiow,ata data. The rnodefing of a nof210 as a pipe in the Case of hsa University to assess the RELAP5/ MOD 2 performance in CFT 24 may give rise of unreasonable results in subcoolod crtt-the application to the natural circulation test A2-77A performed cal flow region.

in LOBt/ MOD 2 facility. Sensitety calculations havs been por.

kwmed in this context, with the attempt to distinDuish the code NUREG/lA 0087: RELAPS/ MOD 2 POST TEST CALCULATION hmitsbons from the uncertainties of the measured cond: tons.

OF THE OECD LOFT EXPERIMENT LP SB 2. PEREZ,J.;

The charactentaton of instabilites in two-pha*e natural carct 4-MEND 12ABAL,R. Consejo de Segundad Nuclear (Spain). Apal tion and the evaluaton of tf o usor effect upon the code results 1992.68pp.0206050073. ICSP.LP.SB-2-R 61893 001, are Special boa!s achiewd in the frame of the A2 77A ana!yMs This document presents the analysis of the OECD LOFT LP-Both of Dese are discussed.

SB 2 Exponmont performed by the Conselo de Seguridad Nu-clear of Spain wo4 king group making use of RELAPS/ MOD 2 in NUREG/lA 0086: ASSESSMEN1 Ot' RELAPS/ MOD 2 CRITICAL the frame of the Spanish LOFT Project. LB-SB-2 expenment FLOW MODEL USING MARVlKEN TEST DATA 15 AND 24 studios the effect of a delayed pump inp in a small break LOCA KIM,K.; KIM H J Korou inW.;te of Nuclear Safety. Apnl 1992.

scenano with a 3anch equivalent diameter break sn the hot leg D6pp 92000$0082. 61893150 of a commeremi PWR.

1 f

,.m..--,-,-

w--

. =.

1 l

Secondary Report Number Index This index lists, in alphabetical order, the performing organization issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross-refererced to the NUREG number for the report and to the 10 digit NRC Document Control System accession number.

i Sf CONDARY REPORT NUMBER Rf PORT NbliOER SECGNDARY PEPORT NUMBER REPORT NUMBER Al.L W M2464 NU4EG/lA 0027 (Cb8^LPJ P-1 NVREG/lA 00$0 Ali W M25%

tlVRl G/lA (p071 iCsp4P FP 2 NUHf G/lA4049 All W M25%

NUHf G/lA4073 IC5P LP S3-1-R NUREG/lA4074 At f W H2435 NUNI G /lA4050 IC5P-L P-58 2 R NUHEG/lA-0067 Al i W-42454 NURI G/lA-0072 Kt i NO 1721 NUALG/CR 5847 Al L W41204%

NUHL G/lA 0079

( A.UR-91 2543 NUHL G /lA-00. e BM62173 NT 163 (90)

NUREG/lA4084 NHHL G/CR 45ff9., VOI N2 OHNL/N36C 200 NURF G/CR-2000 Vit N2 HNL NURf G 52263 W ilG/CR 569 h

(3NL NUREG $2306 (4U41 G/CH 4005 YII N3 DNL NURE G fJ307 NURf G/CH 5306 i

nqqtf 7y q gg,0 NUHf G/Cli 57f55 HNL NUHt'G 'J325 HUHi.G/CH 5875 OHNL/1M 12067 NVAC G/CH4569 HNL,NUHE G 5232H NURF G/CR48 77 OHNt3TM 0593 NUREG/CR 4219 V08 N2 EIN6.-NURF.G 523H NUHf G/CR-5665 OHNI.SUB79 777P8 NURE G/CR4847 EGG 2"06 NUAL G/CH 5604 V01 OHNL $Uus217624 6 NURE G /CA4860 t GG 2600 NUREG/CR %04 V02 Ptn 4711 NUREG/CH A469 V12 f GG 2006 NUREG/CH bW4 V03 PNL 7454 NUREG/CRei33 LOG 2625 NUR[ G/CR-fe66 PNL-7724 NUHF G/CR4821 f GG 2627 NURt G/CH E301 PNL 7776 NURf G/CSU4831 iGGPbE NUHEG/CH 5672 V02 PNL 7092 NURLG/CR 4814 EOG 2643 NUHfG/CR 5720 PNL 7986 NUPEG/CR 5B58 i GG ?b49 NURE G /CH4744 PNL-8051 NUREGICR4870

( GG 2t.50 NUHE G/CR 5745 PNI 8064 NUREG/CH4811 V01 (GG fit?

NUHf G/CH 5445 PWR/HTWG/PA ?562 NUREG/lA-0061 IGG 2666 NUREG/CR 5820 PWR/HTWG/Penc29 NU*h G/lA 0004 I GG 2 tit 1 NUREGiCH $955 WR/HTWG /NB9 NUMG/ LAM EGG 2673 NUnflG/CR 5862 PWRdiTWG/P89 700 NUREG e TAM >$

LGG 2674 NUR(G/CH5865 PWH/PAWG/P68390 NURE G/lA4056 L S TD/t /0117/H09 NURf G/lA-0060 PWRHT WGPte ?)496 NUHEG/iA 0036 SANDB6-1309 NURI G/CR 4%1 V2R1P4 GD/ Pf -N /b9/

NURI G/lA 0063 SANDb8 0906 NUREG/CR-5121 GD/PE N/721 NURE G/lA 0066 SAND 91-2226 NUREG/CH 5/47 3D/PE N/725 NUHE G/lA 0064 ST UDSVIK/NP8763 NUHJG/lA4037 GO'PE -N / 745 NUHF Gila.00%

S1 UDSVikNP68101 NUREG/lA 0038

. IC5P-CO t ur,,W-T NURE G/l A4068 TPRD/t /3209'887 NURE G!lA.0036 ICSP-JC403 R NURf G/lA 0075 1PRDEI S 0154MH7 NUREG/lA 0061 ICLP-JC $PH T quRE G /lA-0062 UCRL-lO 110038 NUREG/CR 5892 17 t

dA A

AA

,*..,4,m,,

mA%-.4-

.-.-a_-_s,

-wm__4ven._4 4

4 4

m.4=,4.._--_,.--_.._~--____--m ee.J,------

--& 4,.4 A4,.4 n

--a.e a.-.w%.A e-A.

en

,44 AaA_a__.mmma_

--.2-._.

a I

=*em a..

s---am_,

m

.m m

.m m.

m

,,,_m.,

-m,,,

p

Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is follsed by the NUREG number and the title of the report (s) prepared by the author. If further information is needed, refer to the main cita-tion by the NUREG number.

AABERGAL BE NSONH AVE R,S.

NUREG/CH $614 EVALUATION OF EWOSURE PATHWAYS TO MAN NUREG/CR 5797 IMPACTS BRC. VERSION 21 Codes And Data Ven4 I ROM DISPOSAL OF H ADiOACilVE M ATE HlAL.S INTO SANIT AriY caton SEWFH SYSTEMS BIRCHLEY,J C.

4D AMS.J.

NUREG/lA 005& RELAP5f MOD 2 ANALYSIS OF Lori EXPERIMENT NURt G/CR 4753 V05 CANADIAN SEISMC AGHE L ME NT Annual L9 3 Repori. July 1969 Jur e 19ido NUREG/lA.00/2. LOrT INPUT DAT ASET REFERENCE DOCUMENT FOR RELAPS VAllDATICN STUDIES AKE RS D.W.

NUMEG/CR 5672 V')2 CHARACTERisitCS OF LOW-LEVEL RAhtOAC-SLAHNIK,DL TIVE WASTE Decontamnahon Waste Prcra% Annual Report For NUREG/CH $670 dESULTS OF LWR SNUDBER AGING HESEARCH F 6 scal Year 1901 BOLLINCE R.O.A.

ALEKANDER.DJ NUREGICR 5731. PIETJMONT SEISMIC REFLECTION STUDY 1 A PRO-NUhEG/CR $785 EXPE RIME NTAL RE SULTS OF TESTS TO INVESil-GRAM INTEGRATED WITH TECTONICS TO PROBE THE CAUSE OF GATE FLAW flEHAvtOR OF MECHANICALLY L )ADED STAINLLSS E ASTERN SEISMICITY.

ETill CLAD PLATES DONENDERGERA AMOS LN-NUREG/CR 584' THE INFLL/ENCE OF PRECOMPnESSION ON THE NUHtG/CR 45$1 VPR1P4 L V ALUATION Of SEVE RE ACCCE NT LOWER DOUND INITIATION TOUGHNESS Or A 533 B RE ACTOR-R!SKS OUAN11HCATION OF MAJOH INPUT PAHAMETERS E mpets' GRADE STEEL Delenanahon Of Gource imm issues ANDRE W,M D NUHEG/CR 4753 VOL, CANADIAN Sf lSMC AGHEE ME NT AnrW Report, July 1969 - Jurie 1990 BOWERMAN.B.

NUREG/CR $885 METALLU4GICAL EVALUATION OF WELD OVER-

gggyy, LAP PIPE SECTIONS FROM BHUNSWICK. UNIT 2 NUCLEAR NUHE G /CF04 753 V05 CANADtAN SE tSM C AGREEMENT Annual POWER STATION Riport. July 1Jti9. June 190.

g DOWA ARCIE RI W,C I4UREG/CH 4551 52h1P4 EVALUATION OF SEVERE ACCIDENT NUHEG/CR 5820 CONSEQUENCES OF THE LOSS OF THE RESIOUAL stSKS OUANTIFICATION OF MAJOR INPUT PARAMETERS Emeris HE AT RE MOv AL SYS' EMS IN PRE SSUR17E D W ATER RE ACTORS Detoranation 01 Source Term issues NUHEG/CR5e5$ THEHMAL-HYDHAULIC PROCESSES DURING RE.

Lt IffVENTORY OPERATtON WITH LOSS OF HESIDUAL HE AT BRANHAM HAAR.K.

NUREG/CR $301: MODELS AND RESULTS DATABASE (MAR-AUFLtCK.J.L DLVEHSION 4 0 Re..erence Manual NUHEGICR 5744 ASSESSMENT Or ISLOCA RISK ME.THODOLOGY BRE G J A D APPt CATION TO A WESTINGHOUSE FOUR-LOOP ICE CON.

RISKS OUANTIFICATION OF MAJC R INPUT PARAMETERS Emperts' NUREG/CR 5745 ' ASSESSME NT OF ISLOCA RtSOMETHODOLOGY Deteirn naten Of Source Term isso -s AND ADPL: CATION TO A COMBUSTIDN ENGINEER!NG PLANT V.

BROOlt.P.

AUSTIN'G/CH $797 NURL IMPACTS ORC.VER$10N 21 Che And Data Voofr NUREG/lA 0065 A

.YSts OF SEM: SCALE TEST S-LH 2 JSING RELAP5/ MOD 2.

caton BAaQ,Y.S.

BROWN.D.P.

NUREGNA-5870 RE SULTS OF LWR SNUBBER AGING RESEARCH.

NUREG/lA M70 ASSESSMf Nt OF RE LAP" / MOD 2 CYCLE 36 04 WITH LOFT LARGE DRE AK LOCE 12 a g

B ARBE R D E.

NUREG/CR 4551 V2R1P4 EV ALU ATION OF SEVERE ACCIDENT NUREG/CR 5677 ASPECTS OF MON!TORING AND OUALITY ASSJR.

RISKS QUANTIFICATION OF MAJOR INPUT PARAMETERS Espe'ts' ANCE FOH RADIOL ABELED ANTIGODtES Detenwnaten On 53urce Term lasues B ARBt HO.F.J.

BRUST,F.

NUHEG/lA 0%0 TRACPF 1 CODE ASSESSP'ENT USING OECD LOFT NUREG/CR-4599 V0i N2 SHORT CRACKS IN PIPtNG AND PIPING LP.FP.1 E XPE RiMENT.

WELDS Semannus Report. Octeor 1990 March 1991.

D E AHM.E.C.

BULLD.R.

NUREG/CR 5732 IODINE CHEM l CAL FORMS IN LWR SEVERE NURE G/l+0064 ANALYSIS OF SEM SCALE TEST $441 UStNG ACODENTS Fir.a1 Report HELAP5/tKJO2 BE AVE RS.J A C AJK A.M.G NUPEG/GR $710 STRESS CORROslON CRACONG STUDIE S ON NUREG/CR 4753 Vos CANADIAN SE ISM C AGREEMENT Annual CANDIDATE CONTAINE R ALLOYS FOR THE TUF F REPOSITORY Report July 1989. June 19%

19

20 Personal Author index C AN T OR.J.A.

DESCHU TT E R.P.

NUhlG/CR t114 HNDINSS OF A WORF SHOP ON DEVE LONNG A NUF/J G/lA D343 ASSE SSVENT STUDY OF RELAPt/ MOD 2 CYCLE ME THODOLOG( F OR EV ALUATIN3 E rF E CTIVENESS OF NLT" E AR 36 D4 B'SE D ON THE ECL L.4 MANUAL LOSS Or LOAD TEST OF POACH PLANT TRAININ3 NDVEUBER 23,19%

CARLSONAW.

DE W ALLKO.

NUHE G/C45%2 A hiGJWAf ACCIDENT INVOLYiNG UNCHADATED NUr,0G/CH 572') MOTOR ONRATED VALVE RESE ARCH UPDATE NUGLE AR F ULL IN SPHINGFIEL D.MASSACHUSE TTS ON DECE M.

OEH 16.1ihel DI AZ. A. A.

NJHEG/CRM71 V01 DEVELC'PMENT OF EOuiPMENT PARAMETER C AR T E R.D-TOLE H ANCE S FOh THE ULTRASONIC INSPECTIO'l OF STEEL NUHEG/C9 $%9 HE ALTH PHYSICS POSriiONS DATA BASE COMPONENTS Applicaten IP Componoots Up To 3 incte ituck CASTRILLOf.

DILLisTONE.M.J.

NUHE GnA Oor,8 ASSESSMt NT OF THE "ONE FEEDWATER PUV" NUMEGUA-0071 ANAL.YSIS Or THE UPTF SE PARATE EF FECTS TEST THIP THANSIENT" IN COF HENTES NUCLE AR POWE R PLANT WITH 11 (STEAM WATE R COUN1dRCURHENT FLOW IN THE BROtEN IU#C # I LOOP H7T LEG) US1NG HELAP5/ MOD 2 DINNEEN'R A NUHf G/lA4145 ASSE S5ML NT Or HELAP5/ MOD 2 USING LOCE NUREG/CR 5301 MODE LS AND RESULTS DATABASE (M AR.

l ANGE BHEAK t.OSS OF. COO (ANT EXF E ft!ML NT L2 5 NUREG/lA 004c AS6f SSME:.1 Or HEL. APSeMOD2 USING SEVIS-DLVCHSION 4 0 Retenence Manual CALE L AHGE UHF M LOSS Or-COULANT L APE HIMENT $463 DOCTOR.S R CH60U.J-L NUREG/Ch 4409 V12 NONDESTRUCTIVE ExAMtNATil NUiH G/nA 0045 ASSESSMENT Of RE LAP 5/ MOD 2 USING LOCE ABillTY FOR INSE RVtCE INSPE CTION OF L IG*

L AnCt EmE AK LO9S 00 COOLANT EXPER!ME NT L2 5 RE AC! OHS Annual Repor10ctober 1989 Septemtiet 19W NUHI G UA M46 ASSESSME NT Or ht LAP 5/ MOD 2 USiNG SEVIS.

NUHE G/CR-%71 V01 DEVI.tOPMENT OF FOUFVENT PARAMETER CA1 E L ANGE DHLAK LOSSOf -COOLANT E APE R:ME NT Sa6 3 TOLERANCES FOR THE ULTRASONIC INSPECTION OF STEEL COMPONENTS Appliceon To Components Up To 3 Inches Thd CHOU.G H.

NUHL Gila 0%0 ASNE SWE NT OF Rt L APS/ MOD 2 USING SE mis.

DR YSU ALE.J.A.

CALE Nit.RME DIAT L bHE AK LOSSOF-COOLANT EXPE htMENT S-NU1E G 'CR -4 763 VDS CANADIAN SEISMIC AGRE E ME N1. Annual 10 ;.

RerotJWy 1989 June 1990 CHUNG.B 0.

Ea.,R R.C.L NUHEG /l A DOW ASSESSMENT OF HELAPS/ MOD 2 CT CL E 36 04 NUHE G/CR 5'10 STRE SS CORHOSiON CRACKING ST UDiE S ON US;NG LOrT IN1E HME DIAT E DRE AK E AF ERMLNT L51 CAND'DATE CONT AINER AELOYS FOR THE 1UF F REPC'SITORY.

CODDING T ON.P.

DW Y E R.P. A.

NUHL G / A4>27; 1RAC-PI 1/ MOD 1 CALCULATIONS Or LOs T EXPLHL NUREG 14% AN AL1ERNAtlVE F ORMAT F OR CATEGORY IFULL l

MENT LP 02 6 C1CLE F ACJY PHYStCAL PROTECTION PLANS CODEtt.A EISENUERO.N.

NUHEG 1327 INITIAL DE MONSTRATION 08 THE U S NRC S CADA.

NUHEG 1327, IN:TI AL DEMONSTRAllON Of THE U $ NRC'S CAPA.

P; LIT Y TO CONDUCT A PErORMANCE ASSESW:ENT FOR A B:U TY TO CONDUCT A PERFORMANCE ASSESSMENT FOR A HIGH LEii L W ASTE REPOSITOHV gigs. LEVEL WASTE REPOSITORY, COOK.K.V'CR 5/B5 E XPEHMENT AL RE SULTS ' r E NCISO,$.

NUF G/

TE SIS TO INVESil.

NUREG /l A 4049 THERMAL HYDRAUUC POST-TEST ANALYSIS OF G ATL F LAW BEHAVtOR OF MECHANICALLY LOADED STAINLE SS DECD LOFT LP FP 2 EiPE4l MENT STER:L CLAD Pt ATES E RiK SSON.J.

CORUH.C.

NUR EG / LA-003 7 ASSESSMENT OF RE LAP 5 / MOD 2 CYCLE 36 04 NJHLG/CH 5731 PIEDMONT SE!SMIC RLFLE CTION STUDY A PRO.

A 5

GRAM INTEGHATED WITH TECTON!CS TO IHOBE THE CAUSE OF EASTERN SEI5MICITY pgggggg,p y, CORWIN,W R NUHEG/lA 0062 ASSLSSMENT OF TRAC F"1/ MODI AGAINST AN IN.

NUHLG/CR 5785 EXPERIMENT AL RESULTS OF TESTS TO INVESTI-ADVERTENT PRES $UR!lER SPRAY TOT AL OPEutNG TRANSIENT GATE FLAW BEHAV6OR Or ME CHANCALl.Y LOADE D STAINLESS IN JOSE CABHL.RA POWER PLANT.

STE EL CLAD PLATES F E HRINGE R.D.

COST AIN.J.K.

NUREG.1327. INITIAL DEMONSTRAT60N OF THE U S NRC'S CAPA-NUREarCR 5731 PdEMONT SEISMIC REFLECTION STUDY. A PRO-OtuTY TO CONDUCT A PERFORMANCE ASSESSU!NT FOR A GH AM INTEGRATED Wi1H TEClON!CS TO PROBE THE CAUSE OF HiGH LEVEL WASTE REPOSITORY E ASTERN SEISMICITY' FISCHER,LE.

CROIFOD.M G.

NUREG/CR %92 A H:GHWAY ACCIDENT INVOLVING U%RRAGATEE, NUREG/iA-0003 RE LAP 5/MDO2 CALCULATION Of OECD LOFT TE ST NUCLEAR FUE L iN SPR!NGFIELD MASSAC%SETTS,0N DECEM L P F W 01.

BER 1619 sit P2/ EOWSKEC.J.

F ORD.W.

NUREG/CR W5 MET ALLURGICAL EVA2 UATION OF WELD Ov0R.

NUREG 132E IN!TIAL DEMONSTRATION OF THE U S NRC'S CAPA-LA?D PIPt SEC110NS FROM BRUNSWi> UN'T 2 NUCL E AR D!UTY TO CONDUCT A PE RFORMANCE ASSESSMENT FOR A POWER ST ATION HIGH4EVEL W ASTC REPOS!TORV.

D AURIA.F.

F R ANCINI.R.

NURL G/LA 0064 RELEVANT ALSULTS OBTAINED iN THE ANALYSIS NUREG/CR 4599 VPt N2 SHORT GRACAS IN PIPING AND P1 PING OF L ODb MOD 2 NATUHAL ClHCULATlON EMAMENT A2 77A WE LDS Semenua! Report. October 1990. March 199 t,

. ALLY.J W.

FULLER.R E.

NURE G %R %47 THE INFLUENCE OF PRECOMPRESSON ON THE NJAEG/CR 5114 FINDINGS OF A WORASHOP ON DEVEi.OPING A LCMR BOUND INfTtATION 100GHNESS Or A 533 B PEACTOR.

METHOOOLOGY FOR EVALUATING EFFECTIVENESS OF NUCLEAR GR ADE STELL POWER PLANT TRA!NWG

~ - - - - - - - - - - - - _

Personal Author index 21 H AGE ME VE R,D-OCCUPATONAL RAD $ATON E APOSURE AT COM.

NURt G 0713 v11 CALASSt.G M.

ME RCI AL NUCL E AR POWER REACTORS AND OTHER PELEVANT RE SrJLTS OBT A'NED IN THE ANALYS'S NURE G'IA 0%4 Or LOW / MOD 2 NATURAL CtRCULAttON L *PER* MENT A2 77A F ACill1if S.19% Twenty Socond Annual Report HALLP C.

A%E h5ME NT of THE "ONE TE EDW ATER PUVP t UhEG4A-0003 RELAP5/ MOD 2 CALCULATON OF OECD LOFT TEST NURE GNA DOM TR.P TRANSTNT" IN COf RL NTES NJCLE AR POWER PLANT W,TH T R AC Elf 1 NUREG!lA.0004 ANALYSIS OF SEMISCALE TFST S LH.1 04NG LP rw.01 RELAPS/ MOO 2 NURE G/tA4065 Af4AD51S OF SEM) SCALE TEST S,LH 2 USif G OALYE AN.W.J.

ASSESSMENT OF 15LOCA RISK - ME THODOL-NUREG/CR 5%4 Vot RELAP$ rMOD2 rGy AND APPLICATION TO A D ABCOCK AND WILCO

• NJCt.E AR POWE R PL ANT Mam Reprxt H ANE Y,LK ASSESSMENT OF isLOCA RISK ME1HODO'.OGY NURE G'CR %34 V02 ASSE 5',ME NT Or ISLOC A N5x. METHODOL-WUREG/CA $744 OGY AND APPLICATON TO A BADCOCA AND WILCO 7 NUCLI AR AND APPtrCATON TO A WESTINGHOUSE FOUkLOOP TE CON.

PO VER Pt ANT Appenece a A H DrusER PLANT.

NURI UCR %D4 V03 ASSE SSMlNT OF ISLCCA RISK. METHODOt..

NgAEG/Ca 5745 ASSESSMENT OF iSiOCA RISK-METHODOLOGY OGY AND APPUCATION TO A B ABUOCk AND WILCOY N !C(E AR AND APPLICATON TO A COM0VSTION ENGtNEERING PLANT.

POWE.R PL ANT Ag.s+n6tes i M H ARLAN C.P.

n NUREG/CR4797. IMPACTS BACNIRSION 21 Codes And Data Ve fu GERTH,0 OF RELAPS/ MOD 2 CYCLE NUREG/LA-0057. ASSES 5ME NT STUDY y 04 BASED ON THE COMViSSONING TEST RE ACTOR TRtP AT um FULt. LOAD AT THE PHtLt.lP59URG 2 NUCLE AR POWE R PL ANT.

HARPERAT.

gun g gf en.45r.1 V2R1P4 EV ALU AT ON OI SEVERE ACCIDE NT N5kS QUANTWICATON OF MAJOR fi4PUT PARAMETERS Emorts' GE RTM AkD.t NUREG/CR4604 V0t ASSESSMENT OF 15tEK;A R)SK, METHODOL-OGY AND APPLICAf TON TO A BABCOCK AND WILCO ( NUCLE AR Deterunat.on C4 Source term Issues POWE R PL ANT Mmn Reput HARWOOD.C.

NUREG/nA 0003 RE t AP5/ MOO 2 CALCULATON OF OECD LOFT T ASSES $ MINT Or ISLOCA R$k METHOD L NURE GICR %04 VQi AND WILGOx NUCLE AR OGY AND APPUCATON TO A BABOOCF (pgw g3 POM R PL ANT Appen6ces A.H NURE G/CR %04 V03 ASSE $5 MENT Of ISLOCA RSK ~ ME THODOL-HL ASLER.P,0.

OGY AND APPLICATON TO A BABCOCK AND WILCOW NUCLE AR NUREG/CR 4469 V12 b ONDESTRUCTIVE E AAMINATION (NDE) R POWE R Pt ANT Amedces 1 M.

ADIUT Y FOR INSE RVICE INSPE CTON Of UGHT W4TER t

t E 4%9 V01 N2 SHORT CHACES IN hPlNG AND PIPINGHE ATH C.

Wi t D5 5cmwnnual Reput. Ottota IPM. March 1% '

NUREG'CR 5820 CONSEQUENCr R OF 1HE LOSS OF THE RESOU I

N MU W

GILLC NUnf GUA4027 TR AC PFt 44001 CALCULATONS Of LorT ExPERI HELTON,J C.

NURE G/CR 4551 V2RtP4 EVALUATON OF SEVERE ACC DENT MENT LP-02 6 SnS. CUANTWICATION Or MAJOR INPUT PARAMETERS Evert OLOVER,L NE DVONT SE tSM:C REF LECTION STUDY. A PRO.

maun Q sowce Mm Isaws NUREGTR $731 GRAM INTEGRATED 'rYlTH TECTON'CS TO PRO 9E THE CAUSE Of HILLR.L E ASYERN SEl$VOTY NURE G'C45B14 EVALUAf TON Or E XPOSURE PATHWAYS TO MA FROM DISPOSAL OF RADtOACTIVE MATERIAL 3 INTO SAfdTARY

coopyg, NUhEG/CR 44M Vt2 NONDESTRUCTIVE E VAMAATON (NDC) REll-SEWER SYST EMS FOR INSE RVICC INSPECT ON Or UOHT WATEA A9:U TY RE ACTORS enual Report.Octote 19M Sepomter SWO HOCKEY,R L NUREG!CA 4469 V12. NONDESTRUCTIVE EXAMINATON ADO RE F Ori INSER ACE INSPECitON OF UGHT WATER ABitrTY RE ACTORS Antnal Report October 1989. Sepemts 1990.

GORE.B1, NUREG/CR '4J3 AUsit ARY F EEDW ATER SYSTEM Rt5LBASED IN.

SPECTION GUIDE F OR THE TURAEY POINT NUCLEAR POWER NUREG/CR EF71 vot-DEVELOPMENT OF EQUIPMENT PARAM TOtERANCES FOR THE ULTRASON:C INSPECTIOf4 Or STEEL AVAIUARY FEE DW ATER SYSTEM NSK BASED IN, COMPONENTS Appncauon To Components Up To 3 inues TNck.

PLANT NUREG'CR %21 SPECTION GUOE FOR THE eEW AUNEE NUCLE AR POWER PLANT MORNO,T.S.

ASSESSMENT Or EELAPS/ MOD 2 US.NG FEpis-NUREGaA4000 CALE INTERMEDI ATE BREAK LOSS OF400LANT EXPERIMENT

GORHAMED, V/R194 E V AL U AT ON OF 6EVERE ACCIDENT NUREG;CR 45M ksnS OUANTWICAllON OF WJOR INF UT PARAMETERS Egerts' IB 3 Detemneon Of Sowce Tem issues h&RSCHELD S.

NUHEG/CR,512t EXPERMENT AL RESULTS FROM PRESSURE TEST-DEVELOPMENT OF EOutMENT PARAMETER ING A 16-SCALE NUCLEAR POWER PLANT CONT AtNMENT.

GRE E N.C.R NUREG!CR tB71 V01 FOR THE Ut.TRA50NiC INSPECTON OF STEEL TOL E R ANCES COVPONENTS ADPkabon To Components Ur To 0 in@es TNck THERJAL-HYDR AUUC PROCESSES DURING RE.

NUREG/CR 5855 DUCEO tNVENTORY OPER ATION WITH LOSS OF HESOUAL GRE E NE,0. A.

NUREG/ CAM 75 E APERVENTAL MODEUNG OF HE AT AND MASS REMOV AL TRAN5fER IN A iM F LU.D DU6t3uNG POOL WiTH APPUCATON TO POLTEN CORE CONCRETE INTERACRONS HYV ARINEN.J.

NUREG4A4059 ASSESSMENT Or RELAP5/ MOD 2 AGAiNST NAT V2R1P4 E V Al t ATION OF SEVE RE ACCIDENT RAL CHOULATION F APE RIMENT S PE RFORMED WiTH THE GREGORY,J1 REWET lit F AC1UTY.

NURE G/CR 4%)

WSKS OUANTW:CATON OF MAJGR INPUT PARAMETERS E nens' Detemmahon CH Scwce fem issues IRWtN.O ft THE 4NrLUENCE OF PRECOMPRESSON ON THE NURE G'CR 4647 LOWER-DOUN9 INrTIATION TOUGHNESS OF A 59 0 REACTO4 OVNTHfR R NUREG/CR %O AGtNG AS*ESSWNT O& COMPONENT COOUNG GRADE STEEL W ATER SYSTEMS IN PRESSU4tED W ATER REACTO%

22 Personal Abihor Index v

ISKANDI R.S K.

KULLJE HG C M.

NUNLO/CH $785 L APf kM N TAL HE bul f S Of TE STS TO IN t bib NUNI G/CH 50% THf MMAL*YDHAUUC PROCESSE G DUH;NG 84E-GAIF F LAW Dl HAVIO% OF blCHANCM t Y LOAM D SI AiNLESS S TE L L Cl AD d 'l Afl S DUCf D INVt NTGM OPlHATION WITH lob 5 OF HE LIDUAL HEAT HEMOg AL J AMISON,J D.

NURI G/CH 48] t S T A T f. OF T H[ mhi W t vACU ATsON 1ME [ Sit KW ALAINE N,0 MATE SIUD'[ S IOH NUCLE AH POAl H PLANT S NUHLGHA-OU82 THE AS$[ % MENT Of HELAPL/ MOD 2 AGANSi lVO JOHNSE N.O W.

gp g q g; 9 NUHt G/CH %% THLHMAL HyDH AUUC PfKcg %t s Dutcrg; pp LAMONTAGNLM.

DUCI D INVL N'OH f OP[ H A TION WI f H 10% Of HE SIDUAl HrAT NUHEG/CH C53 VD5 CANAD:AN 5[ ISMiG AGHE I ML Ni Arnval HI MOV AL HemFtMy 1M9 June 1100 JO W,H N L AN DO W.M.

NUHi G/CH *L$1 V2H tF 4 L V AL U A TION Of LE Vl HE..CDDl NT l

Hfsk.S OUAN t!F ICAflON OF M AJOH INPUI PAR AME TL HS lam 4NUHL G/CH 459p VD1 N2. SHOHT CHACP S IN MNNG AND hMNG M LDS Sermannua Hoport Octat+r 1990. Mpch 1991 (Wemtnabon Of Soutre 7nem Iwoes r

LAPOINTLO.P.

LG CH Mo7 IMPHOvLMENTS IN F1100 OPLHAiED GATE

^

VALVt DE SIGN AND PHLDICilON MODL L S I 04 NUCLE AR POA[ H U

W ANT SYS11 MS bEUH Pwe i f mW HePot1 Septemter 19W Apfd

=

(g g Nuhl G TH W3 AGING ASSE SSMf NT OF COMPONEN T COCF ING K AO.L W All H SYSil MS IN PHf %UH12[D W ATEH HL AC10HU NUHf G/ LAM 45 ASSLSSMfNT Of HE L AP9 MOD 2 USW 3 L OCE l AH'il fMti AK lOSS OF.CO(X ANT f Wt HM( NT L2 5 LEbEA NUHi OHA Oh A%L %Mt NT Of Hi L APUMOD2 USWG SLMts.

NUHf GHAM9 ASSE SSMI N1 OF HE LAP 5/MOU2 CYCLE 30 04

)

CAL E L A%L UNEAK LO5s.or COOL AN T (.WL HME NT S&3 USING LOf T INTI HMLDIATE DHE AK EWEHWLNi Lf 1 Ki f.VILLM B LE E,S.

NUHfG/ LAO M H(( APE MOD 2 ANALYSi$ Of L0l i (>P(H#iNI NUHf GHA OUT7 klLAP5 ASSE SSMEN T ON DIHECT. CONTACT CON.

L9 4 Di NSAllON IN HOnllON'AL COGUHHENT STHAT4 ilD FLOW lt E L L Y,0 L.

NU,H Orch 5t44 A5st %ML N1 Or isLOCA FoM ME THOLO OGY LEHNER.J R HOW rOUH L OOP IUt DON WWEH WS IDENT4 tCAllON AND ASSES 5ME NT Of CONTAW.

MLNT AND HELL ASE MANAGE VENT STHAICG'LS FOH A DWH NUhl GTH 5745 ASSE SSMLNT Of istOCA R1SLME THODOLOGy MAM l' CONT AINV1 NT.

AND APMLAllON 10 A COMuuSilON E NGiNf [ H;NG PL ANT NUHEG'CH 5m APPLICATION Of CONT AINMl'NT AND HElf ASE Kt NNEDY,W E.

M%NAGLVENT STH AT E GIL S TO PWH DH Y-CONT AWME N T PL ANTS NURt G /Cb-5814 E vat UAllON Of EXPO 5UHL PATHWAYS TO MAN i HOM DiSPOSAt OF HADIOACTIVf MATLH;AL S INTO SANitt,HY LIA14 SE nLH 6YS1f MS NUHE G/iA 0000 A55E SSME NT OF HELAPS/ MOD 2 US!NG SEMIS-g pgg LE INTLHMLDIATE DHEAK LOS" Of. COOLANT EWEHrMLNT S.

NUNLG/CH LMa Ht ALTH PHiSICS POS!flONS DATA BAbf Kt HVINT N,7.

LIANG.K S.

NUHl GdA 00M AS$[ $SM! N T Of HEL APk MOU2 AG AWST N AlU NUME G /lA-0045 ASSESSME NT 7f HE LAP 5/ MOD 2 USWG LOCE HAL CIHCUL AllON l %Fl.ONENTS PE Hf (iHMED WI T H THE L ANGE DHE AK L OSS-OFCOOL ANT E WLHfMENT L2 5 Hi aft.tuiACiuly NUHLGHA 0046 ASSESSMENT OF HELAP5/MCD2 US!NG SE MIS.

KILINSKI.T.

C ktE L ARGE BHE Ak LOSS Of-COOLANT EkPf H#lNT SM3 NUHEG/CH 4%9 VOI N2 SHOHT CHAEKS IN P.PNG AND PIPINGLIA01 Y.

WELDS Sermanno# HoporL O(lptw 1MO Me h 1991 NUiM GHA 0045-ASSE SSME NT OF Hf L AP5 /.dOO2 USING LOCE KIM.HJ L AUGL BHEAK LOSS-OF-COOcANT E XPEHMENT L? $

Net G AA-0046 ASSE SSMI NI OF HELAPS/ MOD 2 USWG SE M!S-NUHf G/l A 00M ASSESSMENT 05 HE L AP5 / MOD 2 CEE % 04 U#NG LO81 IN1E HMI DIATE DHr AA E *Pf HWE NT LS-1 cat E LARGE BHEAK LOSS Or. COOL ANT EFE HMENT S 06 3 NUHLG/iA 007D ASbL bSf/ffj! Of it( LA%/ MOD 2 CYCLE M 04 LIN.C.C.

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ASSE S$MLN T Of HELAPN MOD 2 CH! TIC AL FLOW MARK la CONTAINMENT MODLL USING M AHWP EN IEST DAT A 15 AND 24 KIM.K LLOYDR.C.

NUHEG/CH 5821 NUHEG 'IA 00% AhStSSMENT Of REL AP5/M3D2 CfoinCAL FLOW AU Alu AHV F E T.OW AT E R SY5 f E M HISK-BASED IN.

MODEL USWG MAHWklN IEST DAT A 15 AND 24 bPECTION GUIDf f 00 THE KEW AUNEE NUC4 L AH POWU PLANT.

KIM,S.H LOF A RO,R.

NUHfG/lAE?O ASSESSMLNT OF Hl t AP5/ MOD 2 CYCLE 30 04 NUHE 3/CR W3 AGWG AS3ESSMENT OF COMPONENT COOUNG Wi1H L Of f L ANGE [<F E Ak ( OCE L2 3 WATER SYSTE M51N PHESSUHf2ED WATE N HE ACTOIE LU,P. Y.

HL H 5797 IMPACTS DHC VEH5GN 2 i Ccda And Da:a Vmb ca u LYONS,J A KRE SS.T.S.

NUREG 'CH U53 VOS CANADtAN SE!SMIC AGHf E Vf NT An%M NUHt WCH 5?M IODINE CHE %C AL fOHVS IN LW4 Lt vtHE Hepuus W.Jum W AECfDLN15 f erw Heport KRISHN A S WAM Y.P.

NUNE G/CH 451) V01 N2 SHOH f CHACKS IN hP!NG AND hMNG NORE G 1327. INIHAi Of MONsTHATION OF THE U 5 NHC S CAPA thuT) 10 CONDUCT A PEHFOHVANCL ASSESSMENT FOR A WELDS Sormannuat Heport, Octote 1990 Ma@ 1991 HIGH-L EVEL W ASfE Rt POSiTORY

~

k Personal Author index 23 MAHSCHALL.C W.

OLAND.8C.

NUREG/CR-4%9 VD1 N2: SHORT CRACF.S IN PIPING AND PIPING NURE0/CR-5765 EXPERIMENTAL HESULTS OF TESTS TO INVE$TL WE LDS Semiantiuat 84part October two. March IW1.

NURLG'OR 5MO f RACTURE ML>ANOS DASLD F AfLURE ANALY.

GATE FLAW BEHAVOR OF MECHANICAltv LOADED STAINLESS SIS.

STELL CLAD PLATES OLIVE.KL MARTINJ B'R 6814 EVALUATION OF EXPObuRE PATHWA' 1 TO MAN NuRlG/C NUREG 1350 V04 NUCLEAR REGULATORY COMWSSON INFORMA<

rROM DOPOSAL OF RADIDACTIVE MATERIALS INf 0 SAfoTARY TON DIGf ST.1992 Edmon.

SEWER SYSTEMS.

P dMROSE.D.E.

MA200R,TJ.

NUREG/CR $656: THEAMAL-HYDRAULIC PROCESSES DURING RE<

NUREG/ra 5114' FINDING 3 OF A WORKSHOP ON DEVELOPING A DUCED INVENTORY OPERATON WITH LOSS OF RESIDUAL HE AT METHODOLOGY FOR EVALUATING EFFECTIVENESS OF NUCLLAR REMOVAL PuWiR PLANT TRAINING llARKJ.

MCCARTIN.T.

NUREG 1327; INITIAL DFMONSTRATION Or THE U S, NRC'S CAPA.

NUREG 1327.. NIT 6AE DLMONSTRATON OF THE U S NHC'S CAPA-BluTY TO CONDUCT A PERrORMANCE ASSESSMENT FCR A BILITY _TO CONDUCT A PERFORMANCE ASSEbSMLN1 FOR-A HIGH. LEVEL WAS?E REPOSITORf, HIGhlEVEL WASTE REPOSITORY MCCONNELLJ.W

~

NUREG/CR SC/2 V02 CHARAC1 ERISTICS Or LOW >IVEL RADIDAC.

NUREG/CR 6732 IODINE CHEMICAL FORMS IN LWR $EVERE TIVE WASTE Decontaminaleon Weste Program Arvmel Repart For ACOIDE NTS Feal Report feel Year 199t.

PARKHURST.M.A.

MCKAY,ML NUREG/CR.6814. EVALUATON OF EXPUGURE PATHWAYS TO MAN NUREG/CR.M66 FAULT TREE EVENT TRILAND PIPING & INSTRU, FROM DIF.POSAL OF RADIOACTIVE MATERIALS INTO SAN 11 ARY MI NTATON DIAGRAM (F E P) (DITOR$ VE RSON 4 0 Refseonce SEWER SYSTEMS.

P AY NE,A.C.

MCNAMARA.N.

NUREG/CR 4551 V2R1P4, EVALUATON OF SEVERE ACCIDENT NUREG-Oe'J7 Vtf 404 NRC TLD DIRECT RADIATION MONITORING RISKS QUANTIFICAtlON OF MAJOR INPUT PARAMETERS Extsrts'

- Nt TWORK Proutess Horort Octater-Ducomter.1991 tklorrmnation Of Source Term issues NUHf G OH3/ V12 N01: NRC TLD CHRLCf RADIATON MONITORING NETWORK Progress Report January Mafch.1DDE PE N A,JJ.

-l MCNLil.W[Cf04753

~ A N'IPLG VOS CANADIAN SElSMIC AGRE E Mf NT. Annual ReportJuly 1980. Juno 19'#0 pggggggwlg, MINDITABALR.

NUREG/CR-4219 V08 N2. HE AVY SECTON STEEL TECHNOLOGY NURfGilA4074' Rf LAPS / MOD 2 POST. TEST CALCULAT60N or THE PROGRAM Sormannual PrciJtess Repor1 For Ard Septemlet 1991.

OECO LOFT EXPER!Mt NT LP SD 1.

NURt G/lA,0087. RELAPS/ MOD 2 POST. TEST CALCULATON OF THC.

PEREZ).

00C0 LOFT EXPERIMENT LP-SO 2, NUREG/lA.0074: RELAP5/ MOD 2 POST TEST CALCULATION OF THE OECD LOrT EXPERIMENT LP.6itt MILIANL NURE G/It 0087: RLLAP5/ MOD 2 POST. TEST CALCULATION OF 1HE NUREGICR Seat MET ALLURGICAL EVALUATION Or WEL.D OVER<

OECO LOFT EXPERIMENT LP.SB-2 LAID PIPE SFCTONS FROM BRUNSWICK UNIT ? NUCLEAR POWE R ST AlCN PUGHA NUREGICA 5821: AUX 1UARY FLtDWATER SYSTEM RISK 4ASED IN-U G CR 5633 AUX 1UARY FEEDWATER SYSTEM RISK BASED IM.

SPECTION GUIDE FOR THE TURKEY POINT NUCLEAR POWER RADDATZ.C.T.

^

N#

NfE CR 5621: AUXIUARY FEEDWATER SYSTEM Hisk BASED N-SPECTON GUfDE FOR THE KEWAUNEE NUCLE AR POWI A Pt ANT F AuuTIES,19891wenty Socond Annual Report MORCOLN.

NUREG/CR $672 V02 CHARACTEHiSTICS OF LOW.LEVh NDOAC.

R AH M AN.S. -

SHORT CHACKS'IN PIPING AND PIPING TIVE WASTE Decontamirebon Waste fiogram. Annual Repor For NUREG/CR-4599 V01 N2:

FacalYear1991_

WEL DS Senvannual Report. Octoter t wo - March 1991 MUNRO.P.S RANDALL,J.

NURE GICH-4753 VOS CANADIAN SEI M C AGREEMENT Annutt NUHEG-1327; INITtAL DEMONSTRATION OF THE U S. NRC'S CAPA.

AccortJsy 1989. June 1990 BILtTY TO CONDUCT A PERFORMANCE ASS 3SSMENT FOR A gg3 HIGH LEVEL WASTE REPOSITORVJ NUREG1452.

RE VIE W AND EVALUATION RAO.RA TECHNOLOGY EQUIPMENT CODES AND STANDARDS i A 4

NUREG/CR.5791; IMPACTS-BRC. VERSION 21 Codes h1 OMa Verife ZATION OF INDUSTR!AL RADIOGRAPHillLM cat ~n

't NAFF S A NUbiG/CH $855 THERMAL-HYDRAUllC PROCESSES DURiNG RE'-

0,L.

DUCED INVENTORY OPERATION Wf1H IOSS OF RESIDUAL HE AT NUREGAA-0075: FiE LAPS / MOD 2 ANALYSIS OF A POSTULATED AtMOVAL "COID LEG SDLCN SIMULTANEOUS TO A " TOTAL '3LACKOUT" EVENT IN THE JOSE CABRERA NUCLE AR STATON N AV ARRO,A G.

NURE GhA 0068 ASSESSMENT OF THE "ONE FEEDWATER PUMn REVENTOS.F.

TRIP TRANSIENT" IN COrRENTES NUCLEAR POWER DLANT Wif H NUREGAA,0049 THERMAL. HYDRAULIC POST-TEST ANALYSIS OF T M C Brt, OECD LOFT LP-FP 2 EXPERIMENT O'M AHONE VA RHOADS.K.C.

NUREGUA-0073 TME STEP AND MESH SilE DEPENDENCIES IN THE NUREG/CH-5614-. EVALUATON OF EXPOSURE PATHWAYS TO MAN HE AT CONDUCTON sot.UTION OF A SEMMMPUCliflNf70 Dir.-

FROM DISPOSAL Or RADOACTivE MATERIALS INTO SANITARY FERENCE SCHEME FOR TRANSIENT TWO PHASE FLOW -

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4 24 Personal Author tr.dex ROBERTS.T, ST AF FORD.R S.

NURE Q/CR 5885 MET ALLURGICAL FVALUATIQN OF WI LD OV[ R.

NUREG/CR.t%9 HE ALTH PHv51CS POSITONS DAT A P ASE LAO PIPE SECTIONS FROM BRUNSWlCK UNIT 2 NUCLE AR POWER ST ATION STEELE.R.

NURIG/CR 5720 MOTOR OPER/,TED VALVE RESE ARCH UPDATE.

ROBINSON.O.C.

NURLG/CR 5785: EXPT NMENTAL RESULTS OF TESib TO INVEST 6 STEWART.M A.

4 GATE FLAW BEHAVOR Or MECHANICALLY LOADED ST AINLESS NUREG/CR 5KS GENERIC SERVIEf WA1ER SYbfEM ROK BASED 5

STEEL CLAD PL ATES INSPEC10td GUIDE.

ROLLSTIN,JA STRUCkMEYERA NLRLQ/CR4797: IMPACTS-BRCNERSION 21 Codes Ard Data Veir6 NUREGoti37 Vit N04 NRC TLD CaRECT RADIATON MON 110R;NG coton NETWORK Prcgreet Report OctotenDecemt>er.1991.

NUREG-0837 Viz N01: NRC TLD DIRECT RADIATON MONITORING MOSE NTILLO.A R NETWORK Progrett Report January March.1992.

NUREG/GR 5860 FRACTURE MECHANICS BASED FAILURE ANALY-SIS STUBBE.E.J.

NUREGRA-bo43 ASSESSMf NT STUDY 05 RELAP5/ MOD 2 CYCLE DOUEL G.P.

30 04 BASED ON THE DOEb4 MANUAL LOGS OF LOAD TEST OF NUREG/lA-0044 AGSESSMENT STUbY OF RELAP5/ MOD 2 CYCLE NOVE MDE R 23.1985 3605 BASED ON THE TiHANGE.2 REACTOR TRIP OF JANUARY NUREG/tA4044_ ASSESSMENT STUDY OF RELAPS/ MOD 2 CYCLE 11,19 0 3605 BASED ON THE TlHANGE 2 REACTOR Trip OF JANUARY RUSSE LL.K.O.

. NUREG/CR 5301 MODELS AND RE SULTS DAT EA9E (MAR.

SUBUDHl.M.

D).VERSON 4 0 Reference Manual NUREG/CR t@3 AGING ASSESSMENT OF COMPONENT CCOUNG WATER SYSTEMS IN PRESSUHillD WATER REACTORS NUREGIC45720 MOTOR &ERATED VALVE RESEARCH UPDATE.

T A M,P.S.

SCHNURR,N M NUREG4847 SO9 SAFETY EVALUATION REPORT ret 4ED TO THE

^

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NUREG/lA 0078 REVIEW AND

SUMMARY

OF TRAC ASSESSMENT FROM THE INTIHNAflONAL CCDE ASSESSMEN1 AND APPLICA.

Nos $W W 50 m Mmese Va% A#W TION PROGRAM OCAP)

T A YLOP.T.T.

SCHUSTE R.h NUREG/CR-4409 V12: NONDESTRUCTIVE EMMINATION (NDE) REll-NUREG/CR5e85 METALLURGtCAL EVALUATCN OF WELD OVER-ABiUTY FOR INSERVICE INSPECTION OF LIGHT WATER LAfD PIPE SLC1 TONS f ROM BRUNSWICK UNIT 2 NUCLEAH F E ACTORS Annual Report,0ctoter 190 Septemtier 1990

" OWE R ST ATON T HOM A$,J.T.

SCOTT,P.

NUREG/CR-4753 V05 CANADsAN SEISMIC AGREEMENT Annual NUREG/CR-4599 V01 h 1 SHORT CRACKS IN **1 PING AND PIPING Report Juft 1989 June 1990 WE LDS Senuannual Report Octoter 1990 March 1991 THORNE.P.O.

SCRtVE N AH.

NURTGICH 58tu INFORMATION FOR CONSIDERATON IN REVIEW-NUREG/tA-00% ANALYSIS OF Lout TE'iT BLO2 (THREE PERCINT ING GROUNDWATER PROTECTON PLANS FOR URANIUM MILL COL D LEG BRE AK) WITH RELAPS COM TAILAGS SITES a

NUREGilA 0000 APPUCATON OF irsE RELAP5/ MOD 2 CODE TO THE LOFT iESTS L3 5 AND t 3-6 URB ANIK,T.E.

NURLollA 0060 PDL AND POST TEST ANALYSIS OF LOBf MOD 2 NUREG/CR-4R31 STATE OF THE ART IN EVACUAllON TIME ESil.

1EST ST 02 (BT.00) WITH RELAPS/ MOO 1 AND MOD 2 (LOSS OF MATE STUDIES FOR NL* CLEAR POWER PLANT!L FEED W ATER).

VO M SH ANNON,W.E.

NUREG/CR 4469 V12: NONDESTRUCTIVE EXAMINATON (NDE) REll.

NUREG/CR 4753 V05. CANADIAN SEISOC AGREEMENT, Annual ABluTY FOR INSERVICE INSPECTON OF UGHT WATER ReportJuly 1999 June 1MO REACTORS Annual Report Octotsee 1989 Septemt,er 1990 WUREG/CR-5633. AUXILIARY FEE DW ATER SYSTEM RISK-BASED IN.

SIMONEN,FA SPECTION GUIDE FOR THE TURKEY POINT NUCLEAR POWER NUREG/CR4409 V12 N9NDESTRUCTIVE EXAMINATION (NDE) RELL PLANT.

ABluTY FOR INSERVICE INSPECTON OF UGHT WATER -

NUREGICA-5821: AUXILIARY FEEDWATER SYSTEM RISK-BASED IN.

RE ACTORS Annual ReportOctober 1Me. Septen tror 1990.

SPECTON GUCE FOR THE hEWAUNEE NUCLEAR POWER PLANT, SJOBERGA WALKER,C.L NUREG/lA 0038 ASSESSMENT OF TRAC.PF1/MOut AGAINST AN IN.

NUREG/CR-5114 FINDINGS OF A WORKSHOP ON DEVELOPING A ADVERTENT FEEDWATER LINE ISOLATION TRANSIENT IN THE METHOOOLOGY FOR EVALUATING EFFECTIVENESS OF NUCLEAR RINGHALS 4 POWER PLANT.

POWER PLANT TRA!NING SKINNE R,N L W ALT ON.J.C.

NUMEG/CR 5301: MODELS AND HESULTS DATABASE (MAR, NUREG/CR 5445: PERFORMANCE OF INTACT AND PARTIALLY Db D) VERSON 4 0 Faterence Manuat GRADED CONCRETE BARRIERS IN LIMITING MASS TRANSPORT, NUREG/ Cst %66. FAULT TREE, EVENT TREE,AND PIP NG & INSTRV.-

I MENTATON DIAGRAM (FEP) EDITORS. VERSION A J Refere,ce WANG.J K, Manual.

fwAEG/CR $P07; IMPROVFMENTS IN MOTOR OPERATED GATE VAL.

• DESIGN AND PREDICTON MODELS FOR NUCLEAR POWER SMETHURST,A.J.

PLANT SYSTEMS SDIR Phase i Final Report $opteenber 1990 Apnl NUREG/lA.0079 POST, TEST ANALYS4 OF LOBI TEST BT 12 USING 1991.

RELAP5/ MOD 2 W ANG.G-F, SMITH.CL NUREG/f A 0045: A*SLSSMENT OF RFLAP5/ MOD 2 USING LOCE NUREC/CR 5865 GENERfC SERVICE WATER StSTEV RISr< BASED LARGE DREAK LOSS-OF-COOLANT EXPER! MENT L2 5 INSPECTION GUOE.

NUREGhA-0046. ASSESSMENT OF ret 6P5/ MOD 2 USING SEM:$-

gp CALE LARGE RREAK LOSS-OF COOLANi L.;PERMENT S-06-3 NUREG/CR4469 V12-NONDESTRUCTIVE EXAMiNATON (NDC) RFu-WARDLW.

ABluiY ', OR INSERVICE INSPECi10N OF - LIGHT WATER NUREG/C3 5820 CONSEQUENCES OF THE LO3S OF THE RESOUAL hlACTOhS Anna Report.OctoLer 1989. September 1990.

HEAT REMOVAL SYSTEMS IN PRESSUR12ED WATER REACTORS.

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hd.uu. h.

Personal Author index 25 W ATKINS,J C.

WITMILLER.R J.

NURLG/GH $720 MOTO81 OPl4ATI D UAL VE HEL( ARCH UrtATL NUH( G'CH 4763 VD$ CANAD4AN bltbMIC AGHtLMENT Annual RevAJey 1009 June 19i40 WE Bt H.C F.

NUHf GiCH t732 IODINE CHL M; CAL IOHMS W LWH bi VL Hf WILKDWf,Kl.0 M.

ALLIDE NT S f inal Report NUHF " 'C84-4!99 VD1 N2 LHOHT CHACM S IN PIPING AND PlPlN3 W'

s !.e'evawa! heport Odote 199] Match 1991 WEISSAJ.

p NllaiG/CP.0119 V01 PROC (( DINGS Of THE NtNE 'i.E NTH A ATLH gg g

g g

g Hf ACTOH $AF [.T Y INFOf M ATION MF [.1ING

1. 'N # ~

NUHEG/CP 0119 V02 PHOC[ LDiNGS Or 1HL Ni'4f T[f NTH WAf t 4 HC ACT04 $At t TY INrOr4MAloN MI L tit")

WOOD,8 T.

NUHlG/CP 0119 V03 PHOCJE DINGS Of THE NINLiliNi-;l W AT(.R NUHE G/CH 'M F AULT THE F f VENT T At t.AND PiPtNG & INSTHW HLACTOH $Al fT Y IfD OHMATION M[ETING ML N1 ATION DIAGRAM p ( P) (Dif 0HSNE HSiON 4 0 HMru NUHf GICP 0119 V03 AD F'HOCt LDINC.$ Or 1HL NINt it [N1H y,g W All H HL AC10f4 SAf LTY INrOHMAtiON Mt EtiNG V APO.J W.

WEHRY,E V.

NUht G /CR-tBC* api t.lC ATON OF CONT AINML NT AND HELEASE NunLO'CH '870 HL5ULTS Of (WH SNHBBC H AGlNG H[SE AHCH MANAGEMENT S TD *

• E Ull S TO PWR DHY-CONT AINMLN1 WL SLE Y,0.A.

NUHE'G/CH tery honi.'JNiNG MllHODS FOH DEVELOf"N') INT [H-2HANO,X J.

N AL. Nil b$UHf GA8 ACllil S I04 COVDONL N15 IN SYS11 MS NUHf G/CH 5647-THE INC LU( NCE OF F rlECOMPRESSON ON T HE IN1[Hr ACING Wlf H NUCll AH PCWLH M ANf HF ACf 0H COOL-LOWlH-DOUND INIT) alton TOUGHN[SS Or a *'9 0 HEACTOR.

ANT SYSil MS GRAOL STE EL.

m

Subject index This index was developed from keywords and word strings in titles and abstracts. During this development period, there will be some redundancy, which will be removed later when a rea-sonable thesaurus has been developed through experience. Suggestions for improvements are welcome.

A 633 9 Reactor Orode Steel Clad Plate NUREG/C45B47. THr. INru'ENCE Or PRECOMPRESSON ON THE NUREG/CR 5785 EXPERIMENTAL RESULTS OF TESTS TO INVESTI-LOAER BOUND INvilATf0N TOUGHNESS vr A 533 0 RE ACTOR GATE FLAW BEHAVOR OF MECHANICALLY LOADED STAINLESS GRADE STEEL.

STLEl CtAD PLATES ACRS Reports Combust 6on Engineering Plant NUREG-11/5 Vt3 A COMPILATON OF REPORTS OF THE ADV!SORf NUREG/CH-5745 ASSESSMENT OF iSLOCA RISK-METHODOLOGY COMM:TTEE ON REACTOR SArEGUARDST991 Arnval AND APPUCATION TO A COMBUSTON ENGINEERING PLANT.

Abbreviations Component Coolice Water NUREG 0544 f.u3 NRC COL LECTON OF ABBREVIATIONS.

NUREG/CR $$93 AGING ASSESSMENT OF COMPONENT COOLING Abnormal Occurrences NunEG.0090 Via N04 REPORT TO CONoqESS ON ABNORMAL Concrete Darrter OCCURRENCES October Decemtier 1991 NUREU/CR 5445 PERFORMANCE OF INTACT AND PARTIALLY DE.

GRADED CONCRETE BARRIERS IN UMITING MASS TRANSPORT.

NUREG 0544 R03 NRC Cou ECTON OF ABDREVIAflONS Condensation NURE4" LOO 77. RELAP5 ASSESSMENT ON DAECTCONTACT CON-EG/CR993 AGING ASSESSMLNT OF COMPONENT COOLING WATER SVSTEMS IN PRESSURIZED WATER REACTORS Container Alloy NUREG/CR 5710. STRESS-CORROSION-CRACKING 1Tuo!ES ON CANDOATE @WR AMOYS FOR NE W WMM N F /R 1 PIEDMONT SEISMIC REFLECTION STUDY: A Ph0-GRAM iNTE3 RATED WITH TECTONICS TO PROllE THE CAUSE OF Containment EASTE RN 8EISMIQTV NUREQ/CR.5 t21: EXPER) MENTAL RESULTS FROM PRESSURE TEST.

Auxilla9 F-edwater ING A 16 3CALE NUCLEAR POWER PLANT CONT AINMENT.

NURcG/CR 5806. APPUCATKW OF CONTAINMENT AND RELEASE NURFG/CR-5633 AUMLiARY FEEDW ATER SYSTEM RISK DASED IN~

MANAGEMENT STRATEGIES TO PWR DRY CONTAINMENT SPECTON GOCE FOR THE TURKEY POtNT MUCLEAR POWER PL ANT.

peggyg IkREG/CR fs821 AVOLIARY *EEDWATER SYSTEM RISK BASED IN-Contaminat!on Contral SPECTION faulDE FOR THE KEWAUNEE NUCLEAR POWER FLANT-NUREG-1422:

SUMMARY

OF CliERNCB/L FOLLOWUP RESEARCH ~

Averar Dos, ACTIVITiEfL NU4G-ON Vit: OCCUPATONAL RADIATO'l EXPO $URE AT COM-Core Damage MERCIAL NUCLEAR POWER REACi.RS AND OTHER NUREG/CH-5s i V01: ASSET DENT OF ISLOCA RISK - METHODOb l

F ACluTIES 1989 TwenySecond Annual Ropet OGY AND APPUCATION TO A BA9 COCK AND WiLCOX NUCLEAR BWR POWER PLANT Main Report NUREG/CR 5805 OENTIFinA!!ON AND ASSESSMENT OF CONTAIN.

NUREGICA 5604 V02. ASSESSMENT OF iSLCCA RISK. METHODOL.

MENT AND RELEASE MANAGEMENT STRATEGIES FOR A BWR OGY AND APPUCATON TO A BAUCOCK AND WILCOX NUCLEAR MARK N CONTAINMENT' ER RANLAppeWes A H NUREG/CR 5604 v03' ASSESSMENT OF ISLOCA RISK METHODOL-Below R!gulatory Concern OGY AND APPUCATON TO A BABCOCK AND WILCOX NUCLEAR h

G /CR-5797: tMPACTS-BRCA RSION 21.Cocles And Data Vent" NU E / 5 I5 SM NT OF ISLOCA RISK METHODOLOGY AND APPUCAT.ON TO A WESTINGHOUSE FOUR400P (CE CON-Bolling Water Reactor DENSER f 1. ANT.

NUREG/CR 5005: IDENilflCATION AND ASSESSMENT OF CONTAIN.

NUREG/CR 5745: ASSESSMENT OF 'SLOCA RISK METHODOLOGY MENT AND RL.E ASE MANAGEMENT STRATEGIES FOR A BWR AND APPUCATON TO A COMBUSTON ENGINEERING PLANT-M ARh 11 CONT AINMENT.

NUREU/CR 5805 CENTIFICATON AND ASSESSMENT OF CONI AIN-MENT AND RELEASE MANAGEMENT STRATEGIES FOR / RWR Canadian Seismic Agreement MARK lt CONTAINMENT.

NUREG/CR 4753 V05' CANADIAN SEISMsC AGREEMENT. Annual Reporuuly 1989 June 1990.

Coun11rcurrent Flow

. NUREG'lA 0071: ANALYSIS OF THE UPTF SEPARATE EFFECTS TEST Cank 11 (STEAM-WATER COUNTERCURRENT FLOW IN THE BROKEN NUREG-1419 DIRECTORY fy CERTIFICATES OF COMPLLANCE FOR

. LOOP HOT LEG) USING RELAP5/ MOD 2 DRY SPENT FUEL STORAGE CASKS.

Crack Certtilcates Of Comp 4ance NUREG'CR 4599 V01 NT SHORT CT4ACKS IN PtPING AND PlPING NUREG 14HL DIRECTORY OF CERTIFICATES OF COMPUANCE FOR WELDS.Senvannual Report. Octceer 1990 - March 1991, DRY SPENT FUEL STORAGE CASKS Crack Arrest Chernobyt.

NUREG/CR-5785. EXPERIMENTAL RESULTS OF TESTS TO INVESTb NUREG-142?

SUMMARY

OF CHERNCBYL FOLLOWUP RESEARCH GATE FLAW BEHAVOR OF MECHANICALLY LOADED F.TAINLESS ACIIVITIEk STELL CLAD PLATES..

27 e

s

m.... _ _,

28 Subject index NUHE G/GR M47 THE INrt Ut NL ('

OF PHLCOWRF SSON ON THE f e, cuter une isolation t OW! R DOUND IF 'AllON TOUGHNL SS Or A 533 6 hl ACTOH-NURLGdA 0038 A5S[SSMENT 061RAC PF 1/ MOD 1 AGAINST AN %

3 GRADE STLi t.

ADVER1LN1 Fl[DWATIR LINE ISOL ATON THAtJSIENT IN THE FUNGHAL $ 4 POWER PLANf

.4U00G/lA OOM ASSI SWLNI Of Of L_AP5/ MOD 2 CDliiCAL F t OW f eederater Pump MOUL L USING MAHVwlN Tl ST DAT A 15 AND 24 NU4E G /1 A 0068 AS$1SSMENT OF THE "ONE I LLDWATEH PUMP Detonismination NUHEG/CR 1551 V2H1P4 i V ALU ATION OF SEvt AC ACCIDI N1 HISk S QUANTIFICATON 06 M AJOR INPUT PAkAMt TE RS E semts' F6ssion Product Ceteerwebon Of Sourt c 1a.<m Invcs NUHf G/CH 4551 V2R1P4 E V-UATON OF St VI RE A CIDE NT Decontamination Waste Program RIStS VUAN116 ICATON OF seAJOR INPUT PARAMETERS E nuts' NUHI G/CH %72 V02 CHAHACflH:SitCS Or LOA (LVE L HADICAC.

De%nmsawn O Sm Term iswes N EG/CR W IDE N11IICA10N AND ASSE R$ML NT Or CONT AIN llVE W AS11 Decontammaton Waste Pnvam Annaat Repori F o, Fiscal Yea' 1W1 MLNT AND RELI ASE MANAGtMENT STRATEGILS FOR A HWR M ARK ll CONT AINMENT Diagnoshe System NURL G/CH MD6 APPL. CATION OF CONTAtWENT AND HELE ASE NURE G/CH 5720 MOf DH DP! RATED YALVE hl SL AHCH UPDATE MANAGEM mT S1RA T E GIE S 10 PWR DRY.CONTidNMLN!

PL ANTS Ditter6ng Profeestorist Opinion NUHL G 1414 DJ i LRiNG PHOF LS90NAL Vlf WS OR OMNtONS 1W3 Flow Behavior Spoual Heww hitel NUHt G/CH 5705 t APEHIME NT AL RISULTS OF 1ES1S 10 INVESil-GA1E FL AW BlHAVIOR Of MECHNICALLY LOADEU Sf AINLf SS Digitization sit t t ct49 pt A1[ S NUHL G 1452 NE VIL W AND EV ALU ATION Of 1[CHNOLOGv.EOU!PME N1 CODE S AND ST ANDARDS f OH ()lGlil-Fracture Mechanics IAllON Or INDUSTH Al HADiOOHAPHIC Fit M NUREG/CH 4460 V12 NOND[STRUCTIVE ExAMINAllON (NDf) DEL 1 Auf tlf Y 108 INSE RVICE INSPE CTION OF LIGHT WATE R

[4r

[

C ACkS P NG ND PSING NU 09 0 NHC COMPREHE N%E DC COUDS DISPO9ttlON y

V

)

W DUlf WILDS Semsannual Hoport Octoter 1990 Mwch 1991 NURE G/CR 5860 F RACTUR( MECHAN CS DASED f ALLURE AN ALY.

Dosage CaHbration NUH( G/CH 58/7 A"P[ CTS OF MONITORING AND OUALIIY ASS' b3 ANCE F OR H ADIOLAllLLE D ANilDOD'E S Fracture soughness Eastern Salsemcity NUREG/CR 4219 VD0 N2 L /Y SECTION STELL 1ECHNOLOGY NURE G/CH $731 Ptl DM?NT SCISMiG R[I t( CTION STUDY. A PRO.

PROGR AM Somiannual Progress Report F or Apni September 1991.

GRAM INi[GRAfl D WITH TECTON:CS TO PHOBE THE CAUSE OF NUHEG/CR 5785 EXPEHIME NTAL HESULTS OF TE STS 10 lNVE ST6-E AS11 RN SEISMICIT Y.

GATE FL AW UEHAVOR OF MICHANICALLY LOADED STAINLE SS ST E t t CL AD PLATED Emergency Planning NUREG/CH 5860 iRACTURL MECHANICS BASED F AILUHE ANALY-NuitE G /CR,4831 SIAT[ OF THE AR1 IN I VACUATON 11ME E STI.

SIS M ATE STUDiL S FOR NUCil AR POWER PL ANTS Fuel Accident Emergent;y Response NUREG 1458 EMERGENCY RESPONSE TO A HIGHA AY ACCIDENT NUuf G 1450 E Mt RGt NCY Nt'SPONst 10 A HIGHWAY ACCIDE NT IN SPRINGFILLD, MASSACHUSETTb ON DECEMBER IfL1991.

IN SPRINGf IEt D M ASSAC HUM ITS ON DECt:MDF R 10.1991 NUREG/GR 5BJ2-A HIGHW AY ACCIDf NT INVOLVING UNHRADIA1E D NUHI GiCR be92 A HIGHA Av ACCOENT INVOLVING UNWHADIATED HUClEAR IUEL IN SPRINGFLELD. MASSACHUSETTS.ON DECEM-NUCLE AR I UE L IN SPHiNGF IE t u MA".SALHUSETTS,0N DECE M-DER 16,1W1 DE H 1E1991 F ull Ecad

)

Enforcement Act6on NUREG/lA4)057. ASSE SSME NT STUDY OF RELAPS/ MOD 2 CYCLE NUut G-0943 V11 N0i ( NFOHCEMLNT ACTIONS SIGNiflCANT AC.

36 04 BASED C'N THE COMM SSIONING lEST REACTOR THIP AT llONS RE SOLVI O Quartedy Progens ReportJanuary Mach W FULL LOAD AT THE PH LLIPSBUHG 2 NUCLE AR POWER PLANT.

Evmiahon

^U" U

URE CR 5720 MOTOR OPERATED V ALVE RESE ARCH UPDATE Al S D $ r NU C aft UP S

NUREG'CR4M7-IMPROVEMENTS IN MOTOR OPERATED GATE Event Tree VALVE DE SIGN AND PREDICTION MODELS F OR NUCLEAR POWER NUREG/CR S6M F AULT THEF [ VENT TRE E.AND PIPING & INSTRU.

PLANT SYSTEMS SBIR Phano i Finat Report $eptemLmr IP7} April E NT ATION DIAGRAM (Ft P) [ Dl10HS VE RSON 4 0 Rotercree IM Groundwe*er Euperiment S.lB-3 NURE G!CR 5445 PEHFORMANCE OF INT ACT AND PARTIALLY DE.

NURE G/iA 0080 ASSESSMENT OF REL AP5/ MOD 2 USING SEMIS.

GRADED CONCRETE B ARRIERS IN LIMITING M ASS TRANSPORT Calf INTEHMEDI ATE DREAK LOSSOF CCOLAN1 EXPERIMENT 5 NUREG/CR 585H. INFORMAllON F OR CONSIDERAT6ON IN REVIEW-10 3 ING GROUNDWATER PROTECT 60N PLANS FOR VRANfuM MILL T AIUNGS Si1ES Euposure Pathway NUREG/CR 5814 EV ALUATON OF E XPOSURE PATHA AYS TO MAN Healm Ptiys6ce IHOM D!SPOSAL OF RADCAC1tVE VATE RIALS ANTO SAN,f ARY NUREG/CR4569 HE ALTH PHYSICS POSITONS DATA BASE.

St WEH StSit MS Heat Conduct 6 ort Fault Tree NUREGilA-007311ME STEP AND MESH S:ZE DEPENDLNCIES IN THE NUREG/CR $6M F AULT THil1 EVENT TREE.AND PtPING & INSTRU-HE AT CONDUCTON SOLUllON OF A SEMblMPLICITJINITE D4F-MENTA10N DIAGRAM tr E P) EDITORS VE RSON 4 O Refe'ence F ERENCE SCHEME FOR 1RANSIENT TWO-PHASE F LOW.

Meual Heat Transter F eedwater NUREG/CP 0119 Vot: PROCEEDINGS OF THE NtNETEENTH WAT[R NURf G/lA-0061 PRE-AND POST.it ST ANALYPS OI t081 MOD 2 REACTOR SAFETY INFORMATION MEETING IEST S102 (BT on) wtTH REL AP5/ MOD 1 AND MOD 2 (LOSS OF NUREG/CRO119 VO2 PROCEEDINGS 05 THE N NETEENTH WATER f EED W Alt H)

REAC10A SAFETY INFORMATON MEETlNG,

l Subject Index 29 I

NUH[G/CP 0119 V03 PHOCllD!NGS OF THE NiNtit f. NTH W Af t.H inadvertent heasurtser sprey fil ACTOH MAS t if Nf OHMAllON Mii11NG NUHf G/IA (M2 Aht! SSME N1 Of THAC Pt 1/MODt AGAINST AN IN NUHI G/CP 0119 W3 AD PitOCf L Dl4G5 Of THE NINLi(INTH ADVLHf(N1 Phl b5Uldlt h $PH AY TOT AL OPEN!N31RANS'I NT W AllH HE ACTOH 6AlLTY INF OHMAllON M[EllNG IN JO$E CABHf HA POWL H PL ANT Heavy Sect 6on $ teel Technology Program industrial Radiographic Film NUHf G/Ui 4219 V0H N2 Ut AVV41 CTION LT([L 1(CHNOLOGY NUHE G 1452 Of Vlf W AND E V Al UAT ION U"

PHOGHAM bonuanrwat Pr97ent flgnr1iof Apfd Septemime IW1 1[CHNOLOGY.EOUIPMINI CODIb AND $1 ANDAHDS f OP DIGif Hi{phlevel Wnste D6tposal 2 ATION Or INDUb f Hl Al H ADIOGH# PHIC ilt M NUHf. G/CH 5 710 Siht SS COHHO510N-CHAC* LNG ST UD'l % ON Industrial Hoom 'aphy CANDIDA s t CONI A!NLfi Allo (S I OH THE f ui t HLPO5110HY.

NUHl G-0713 y11 OCCUPAllONAL HAD:ATION EWO5Ulc Al COM-H6gh-levet Wasto ftepository ME RCIAL NUCLE AR POWTH HE ACTOHS AND 01HLh NUHLG 1327 INtTIAL Di MONS 1HAik3N Of THE U S NHC S CAPA-f AClLlilES,1% lwenty-hetorvj Annual Hepyt Bil N Y 10 CONDUCT A PEHiOHMANCf ASSI$LMtNT fOH A H60H LI VI L WASTE Hi POh!TOHe informatton Circular NUlit G 0725 fl0P PUHL IC INF OHM AllON ClHCUL AH TUH fMIP-Hot leg MINTS OF IHHADIAlf D HC AC10H f UCL-NUni GaA 0071 ANALY51S Or THE UPfl Li PARATE [f f l C% TIST 11 (STI AM WAf LH COUNi[ HCUHHIN1 f LOW IN TH( DHDkl N informauon Digest

~-~

t.004 HOT t t G) USING ht l APS/ MOD 2 NUHE G 13$0 V04 NUCLEAH ft[ GULAlOHY COMMIS$10N IN00HMA-llON DIGLbf 1992 Emtion gpp NUHL GUA Of 0f; ANALY5tS Of 1004 f( L1 ULO2 (f HH(( P(ItCE NI inspection Oude NUHfG/

D AUmIUAHY H I DWA10H SYsit M HISCHASt D IN-N t /

7 t

)

i L AP5/ MOD?.CYClt 36 04 WCilON OULDL f OH THE TUHKIy POIN1 NUCtI AH POWtk Ar A%T t Of f SMAtl HHI AK ( xP('HiMi Ni t 3 6 N'lHI GNA 0041 A%I SLMI N1 $1UDY 06 HLL APL/ MOO 2 CYCt L IIANI 4 04 DAblO ON 1Hi dot t 4 MANUAL LOSh Of 1OAD lt 51 Of NUHLG/CH M21 AuxlLIAHY f[LOWAllH SYST(M h A BASED IN.

NOV[ Mtil H 21,19%

WE CTION GUIDC iOH THE FfW AUNEI: NUCli AH PNTH PLANI.

NUHl G/lA 0044 A%L 55MI NY STUDf Of HELAP5/ MOD 2 CYCtt NUHLG/Ut WM Gt N[ HIC L( HVal W A1( H SYSit M HI5K-BAEED 30 Ot HALI D ON THf ilHANGE 2 HL ACTOH 1 HIP OF JANUAHy tNSPECTION GUIDE 11 1 WI NUhl Gila 0045. Ah5E SSML Ni Of HE t APS/ MOD 2 U5ING L OCE todine t AHG[ Fluf AK LOSS Of COOLANT I RP! H:MI Ni t 2 $

NUHi G<CH $732 IODINE CHt L'lCAL iOHMS IN LWH SLVIHL NUH( h4A 004h Abbt % MtNi Of HL i APL/MDO2 UStNG blMts ACCIDi N1S f mal Heport CA!t l AHGI Of tl AK LO% Or COOL ANT I XPI HiMt NT S 0fa 'l 2

NUNI briA 0049 IHf HMAl-HYDHAULIC POSI TEbi ANALYLIS Of LEH Of CD i Of 1 l P f P 2 i &l HIMt NT NUHLG/CH 2000 V11 N2 UCENSE L i Vi N1 filPOHT (t E H)

NUH[ Gila 060 1HACPr 10% ASS (.SSMIN f USING OLCD LOf 1 COMPILAllON f ur Mouh Of f obruary 1Fl2 t P f P 1 ( YPI TUMt NT NUHE G/CH1000 Vit fu tICLNSLL E V1 N1 fit POHT (Lt HI NUH( G4A 007 t ANAL VSIS Of THE UPil RFPAHATL i fi f CTS Tf St COMPILAllON For Vonth Of March 1992 11 IStf AM WAIiH COUNTt HOUHHf Ni ILOW IN 1HL BHOFLN tOOP HOf I[G) USINO Hf 1 APS/ MOD?

LOB 1 F acihty NUHi G/lA 0074 Hf L APS/ MOD 2 POS1.il $1 CALCut.ATION Of THE 2

NUHLG!!A 0061 PHL-AND POST li Si ANAL YSIS OF L001 MOD 2 NU L 4 7

%i

/l )

AN AL YSIS Of A POS1ULAIL D IIU

" COL D tl G SUL OCA" LIMut T ANLOUS 10 A "101 AL Ut ACK OUY" i VI NT IN 1HL JO5l CAliHf HA NtK;tI AH S1 AllON LODI Test NUHl Gil A D)7ts HLVIE W AND SUMMAHY Of THAC ABS! 55Mf NT f HOM THE INT [HNAllONAL CODt ASSE LSML NT AND APPUCA-NUHt G/lA-0079 POST 1E ST ANALYS:S Of LODI it si ut-12 U$iNG I

flON PHOGH AM (ICAP)

M l#

UU2 AMP ACT S-DHC LD01 Test BLO2 NUHLG/Q %# 7 IMPACTS BHC,VI HSiON 21 Coeles And Dma Vmfr NUHI Gil A Om6 ANAt YSIS 06 LODI TE ST DLO2 (THHf f PlHCLNT t atm COLD t.EG DHL AK) WITH Hf LAPS CODL IRHAS LOENMOD2 NUHL G / CH. %M f AUL f THL t TVE N1 101 t,AND PtflNG & INS 1HU.

NUH0 G HA-0064 H[L EVANT Hi SUL.TS OUTAIN[ D IN THf ANALYSIS MLN1 AllON DtAGHAM (f(P) ( Dif OHS V1 HSION 40HWernS OF LOOL / MOD 2 NATUHAL CIRCULAtlON E APTHiMLN1 AT77A Mar e tOCA ISLOCA NURt G/CH 962 SCHI(NiNG ME THODS TOh DEVLtOPING IN11f4 NUHLG/CH 5rsO4 V01 ALSiSSMENT Of ist OCA HISE W THODOL NAL. PHESSURE CAPAClflLS F OH COMPONLNTS IN SYSTEMS OJf AND APPUCATION TO A DAHCOCK AND WILCO 4 NUCLE AN INT [HF ACIN3 Wl1H NUCl[ AH POWl H PL ANI HL ACTOR CCOL-POWI H Pt ANT Mam Herut ANT SYStLMS NUH( G/GH%04 V02 Abbt SSMENT Of ISLOCA HiSK - MC1HODOL.

OGY ANG 1"M 6 CATION 10 A DAHCOCA AND Wit COr NUCt f AR LOCE POWL H PLANT Apowkes A H NUHLGHA 004% AS$fSSMENT Of HC L APS/ MOD 2 USING l OCE NUhtG/CH %04 V03 Abst S5MI NT Of t$LOCA Hi%K - ML1HODOL' WM Wh WS$ Of gum DPLHiMLNT L2 5 OGY AND APPUCAtlON 10 A H ABCOCK AND WILCOX NUCL E AR POWI H PLANT Aptw&cu 1 M LOCE t.2-3 NUHf G/CH 6744 AL5t S5Mf Ni Or ISLOCA HISK Mf 1HODOL OGY NUNE GnA.on70 ASSE SSME N T Of HI LAPS / MOD 2 CYCLE 36 04 AND APPLICATION TO A wt $1tNG 60VSE FOUH tOOP ICE CON' WiiH !Of f L ARGE HHE AK LOct LP 3 Di N%f H Pt ANT NUHi U/CH 5745 A55LSSMLNi Of thtOCA HGk ML THODOtOGY WT AND APPt ICATON iO A COMHUSTION INGINI( HING Pt ANI WHE G AA 00$0 TH AC PF1 CODE ASSE SbMC NI USING OECD LOf1 IVO Loop Sest LPJ P 1 L APE RMLN1 NUR[G4A 0062 TH[ ASSL SSM[NT Of Hf ( APL/ MOD 2 AGAIN$i IVO NUHE G!lA 0072 iOf f INPUT DAT ASL1 HEi f HE NCE DOCUML NT I OOP SF AL T E ST S IOH HE LAP 5 VAUDATION STUDIES NUHf GAA 0075 fil L AP5/ MOD 2 ANALYSIS Of A POSTULATED Ice Condenser

" COLD t EG EBLOCA" SIMULT ANIOUS TO A "TOfil BLACK OJT" NUHi G/CH 5744 A%[ SSME N T Of ISLCK A H15* Mt 1HODOLOGY t Vf NT IN THL JOSE CAHHf HA NUCLf AH ST AllON AND APPLICAilON TO A W1 StiNalOUSL TOUR LOOP 101 CON NUAEG/tA 006E HEL APS/ MOO 2 POST TE ST CAL CUL A' TON Or THE Dt NSt H PL AN1 Of CD LOf1 LXPtH;ME NT LP 5112

l 30 Subject inoex L OF f E s peHnient Loss Of Cooient Accident NU4( G /lA@74 hf.L AP5/ MOD 2 POS1-1[ST CALCULATIDN OF TH[

NUPEG4A 044 ANALYSIS OF SE M: SCALE TEL S-L H 1 USING Ot.CD LOF1 E xiT NMENf 1P SD 1 RE L APS/ MOD 2 NUREG4AW5 ANALYSW OF SEM: SCALE Test S LH 2 UstNG LOFT Esperiment LD 3 R E LAP 5 / MOD 2.

NUFR G aA M LB HE L APL/MDD2 ANALYSIS OF LOf1 ESPE R1MLNT (9 i Low level Radioantve Weste NUREG'CR 5072 V02 CHARACTsR'STICS Or LOW LEVEL RADIDAC-royasAnnual Report For LOFT E speriment 8 6-4 Tid AASTE Ewon'anwnahon Waste a

NUhtGNA& M M LAPL/ MOD 2 ANAL YSIS OF L OF T l KPt H MENT F x aitear19irl L44 Low 4svel Weste Disposal LOF1 E aperiment LP 024 N HEG'CR4797 IMPACTS BRC. VERSION 21 Codes And Date Vehfe NUf 4( G4A OC27 TRAC PF 1/ MOD 1 CALCULAflONS OF LOFT E&EW-caton MENT LP.02 0 MAR D LOFT Intermediate D.pak NUREG/CR 5301 MODE LS AND HESUL15 DAT ABASE (M AR-NURE G/lA D%9 A%E SSME NT OF-RE L APS / MOD 2 CYCLt 3004 D3 VERSION 4 0 Referente Manua:

U5iN3 LOFT INTE HML D.A1E 04f Ak ( a f t RiMLNT LE1 Mark 11 Contsinment LOFT Large Dreak NUREG/ CAM 05 IDENTIFICATION AND ASSE% MENT OF CONT AIN NURE GUA-O' 70 A%E MMENT OF Af L AP5/ MOD 2 CYCLF % D4 MENT AND RE LE ASE MANAGEMr di STRA1(GIES FOR A DWR Wi1H LOF 1 L A4GE but AK LOCE L2-3 MAR A Il CONT AINVE N T LOFT Test Marviken Test Data NUht G U A,0060 AFPLICATION OF THE RE LAP 5/u702 CODE TO NUREG'IA>0066. ASSE SSMENT Or htLAP5/ MOD 2 CRitirAL F LOW 1HE TOFT f[ STS L3 5 ANU 13 6 MODE L USING MARVWEN fEST DATA 15 AND 24 LP f P-2 E speriment Mass Transport nun [ G AA 0049 THERMAL-HYDRAULIC POST t[ Si AN AL YNS OF NuREGiCR +445 %.RFORMANCE OF IN1 ACT AND PARTIALLY DE.

OECU LOF T t P f P P E yPt WMt NT GH ADE D CONCREt E BARR$ RB IN LIMITING MASS TRANSPORT LWR Mesh Sire NUREG/CR 44M V12 NONDE $151UCilvE E KAMWATION (NDE) RELi-NUREGaA.0072 TtME $TEP AND MESH Sl/E DEPENDENCIES IN THE AD4lTY FOH INSE RVICE INSPE CllON OF UGHT W ATE R HE AT CONDUCTION SOLUTION OF A SEMMMPLICIT FINITE Dif-HE ACT ORS Annual Aerort O( tcL,= 1999 Septemte wo FERENCE SCHEME FOR TRANSlENT TWO PHASE F(OW NUREG/CR 5732 IODINE CHE M!C AL FOHMS IN LWH SEVERE ACCID ( N i S F inal Repori Metallurgical NUHlG/ CHM 70 HFSULTS OF LWH SNUUDE R AGING RESE AOCH NUREG/CR M85 METAt LURGICAL EVALUATION OF WELD OVER-LAID P!PE SE CTIONS FROM BRUNSWICis UNIT 2 NUCLEAR Large Break PO AE R ST ATION NUME G/lA 0027 TiinC PF 1/MODt CALCULATIONS OF LOF i i APERI-MfNTLPC20 Models And Results Database NUHEGICR 5301. MODE LS AND RE SUL T S DATA W E (MAR.

Legal lacuences DLVIRSION 4 0 Reference Manual NUHEd O750 V34102 INDEsES TO NUCt f AR REGUL ATONY COM MISSION ISSUANCE $ J/v Lbcomte 1941 Molten Core Concreto NURE G-0750 V35 tot INDEkES TO NUCLEAR REGULATOnY COM-NUHECeCR 5675 E APEPtMENTAL MODELING Or HEAT AND MASS M!SNION ISSUANCE S January March 1992 THANSF ER IN A TWO-FLUID DVDBUNG POOL Wi1H APPUCATION NUHE G 0750 V35 NO2 NUGt E AR RE GUL ATORY COMMISSION IS-70 MOtTEN COHE-CONCRETE INTER ACTIONS.

SUANCf S F 00 FE BRUARY 199? Pays 47-ft2 NUHE G 0750 V35 ND3 NUCLE AR Rt GUL A1OR; COMMISSION IS-Motor Opvated Valve GUANCE S i 04 M AHCH iW2 Pays 83144 NUREG/CR 5720 MOTOR-OPERATED VALVE REC' ARCH UPDATE NUHEG 0/50 VJ5 NO4 NUCLE AH RE GULA;OHV COMMISSION !S, SUANCE S F04 APHIL 1992 Page 1451B1 Natural Circulation NUREG/lA 0059 ASSESSMENT OF RELAP5/ MOD 2 AGAINST NATU-Licensee Event Report RAL CIRCULATION E X PE RIMdNTS PERFORMED WIT H THE NUHIG/CH 2000 V11 N2 LICE NSE E EVENT REPORT lLEH)

HEWET >lll F ACIUTY.

COMPIL ATION For Monm Of Fetirva y 1D92 NUHEGilA-0084 RELEVANT RESULTS OOT AINED IN THE ANALYSIS NUREG/CH 2000 V11 N3 LICE NSE E EVENT REPORT (LE R)

OF LODl/ MOD 2 N ATURAL C!RCULATION EXPERrMENT A2 77A COMPILATION f or Month 04 March 1102 Nondestructive Evaluation Light Water Reactor NUREG/CR 5871 V01: DEVELOPMENT OF EQUIPMENT PARAMETER NUREG!CR 44fP V12 NONDE STRUCTIVE EXAM l NATION (NDE) REll-TOLERANCES FOR THE ULTRASONIC INSPECTION OF STEEL ADiLITY FOh INSE RV4CE INSPE C TION OF LIGHT WATER COVPONENTS Appitcation To Co.nponents Up To 3 Whes Thick.

hE ACTORS Annual Report Oct&ar 19tG Septemter 1/ ^

NUREG/CR 57R IODtNE CHE MICAL FORMS IN Lwn SEVERL Nondestruttive Emamination ACCIDENTS Final Report NUREG/CR 44M V12 NONDESTRUCTIVE EXAM: NATION (NDE) REL6 NUREG/CRM70 RESULTS OF LWR SNUDBER AGING RESE ARCH ADIUTY FOR INSE RVlCE INSPECilON OF UGHT WATER RE ACTORS Annual Report Octoter 1989 September 190 NUREG4A0 W HELAP5/MOC2 ANALYSIS OF LOFT E APE R! MENT Nuclear Regulatory Research LB 4.

NUREG 1266 VM NRC SAFETY RESEARCH IN SUPPORT OF REGU-Lots 07 Load NURE GU A 0043 ASSE SSME NT STUDY OF RELAP5' MOD 2 CYCLE Nuclear Terms 36 04 B ASED ON THE DOEv4 MANcAL LOSS OF LOAD TEST OF NUREG.0544 R03 NRC COLLECTION OF ABOREvlATIONS NOVf MDER n 1995 OECO Loss-Of 4 oolant NUREG/LA.0050 TR AC-PF1 CODE ASSESSVENT USING OECD LOFT NURE GnA 0045 ASSESSMENT OF RELAP5/ MOD 2 U$tNG LOCE LPTP 1 EXPERiME NT.

LARGE BRE AK LOSS-OF-COOLANT dPERIMENT L2 5 NURE GIIAOD74 RELAP5/ MODE POST TEST CALCULATION OF THE NUREGnA 40 ASSFSSMWT Of PELAPS/ MOD 2 US:NG SE MtS-DECD LOFT E XPER> MENT LP-SO-1 CALE INTERMEPATE DRE AA LOSSUF. COOLANT EXPERMENT S-NUREG 4A-0007 RE! AP5/ MOD 2 POST TEST CALCULATION OF THE 18 3 OECJ LOFT EXPER: MENT LP-SD 2

Subject index 31 OECD LOFT Test LP4W 01 NUHf G/CR $121: EXPERIMENT AL RESULTS f HE4 PRESSURE TEST.

NURE G AA4063. HEL APS/ MOD 2 cat CULATION OF OECD LOF1 1EST ING A 10 SCALE NUCLEAR POWER PLANT CONTAINVENT.

LP4W D1 Presourlaed Wetet Reactor Office 01 The Inspector General NURE G/CRM93 AGING ASSESSMENT OF COMPONENT COOUNG NUREG 1415 V04 NO2.

OF FICE OF THE Ifd5PE CTOR WATER SYSTEMS IN PRESSURIIED WATER REACTORS GENERAL Semaannual F4eport Octater 1,1991 March 31.1992 NURE WCR4806 APPLICATION OF CONTAINMENT AND RE. LEASE M AN AGE ME NT STRATEGIES TO PWR DRY CONT AINMENT Of f 6ctal Record pt Aytg.

NUREG-0910 H02 NRC COMPRt;10NSIVE Rt CORDS DISPOSITION NUREG/CF14620 CONSEOVENCES OF THE LOSS OF THE RESIDUAL c

SCHE DULE HEAT REMOVAL SYS1 EMS IN PRES $URl2ED WATER REACTORS.

PRA Probabilistic Risk Assessment NUREG/CR-6301: MOOLLS AND - RESUL1S DATABASE (MAR-NUREG 132L INITIAL DEMONSTRATON OF iME U.S NRC'S CAPA.

DI VERSION 4 0 Re'ereme Manual-B UTY TO CONDUCT A PEHrORMANCE ASSEstMENT FOR A PWft HiGH. LEVEL WASTE REPOSITORY.

NUREG/CH M93 AGING ASSESSMENT Or COMPONENT COOUNG NUREG/CR 4551 V2H1P4 EVALUATON OF SEVERE ACCOENT W A1ER SYSTEMS IN PRESSUniff D WATE R RE ACTORS HISKS. QUAN RFICATof' Or M AJOR INPUT PARAMETERS Experts NUREG/CH $606 APPUCATION OF CONTAINMLNT AND RELE/*J Detenmaton Of Source Term lasues MANAGLMINT STRATEGIES TO PWR r d.CNT AINMENT NUREGICR $004 V00 ASSES $ VENT or ISLOCA RISK. METHODOL.

PLANTS OGY AND ANATON m A BABCOCK AND W@ NWAN.

NUREG/CR 6820 CONSEQUENC'S OF THE LOSF Or THE RESIDUAL POWER PLANT Man Report HEAT REMOVAL SYSTf MS IN PRES $UHi2ED WATin REAC10RS NUREGiCR 5004 V02 ASSESSMENT OF ISLOCA RISK. METHODOL.

OGY AND APPLICATION TO A BABCOCK AND WILCOx NUCLEA9 Peak Cladding Temperature POWE R PLANT Apperdcco A H.

NUREG/LA 0009 ASSE SSMENT OF RELAPt/ MOD 2 CYCLE 36 04 NUHEGICR 5004 V03 ASSESSMENT OF GLOCA RISK + METHODOL.

USING LOrT INTERMEDIATE BRE AK E XPERlVENT L$4 OGY AND APPLICATION 10 A BABCOCK AND WILCOX NUCLEAR POWER PLAlit.Appereces t-M Petittone f or Rulemaking NUREGICRM33 AUXIUARY FEIDWATER SYSTEM RISK BASED IN-NUREG DD36 Vit Not NRC RE GULATORY AGE NDA Ouariony SPECTION GUIDE FOR THE TURKEY POINT NUCLEAR POWER HeeportJanuary Marcti 'W2 PLANT NUREG/CR !J44 ASSESSMENT OF ISLOCA RISK. METHODOLOGY Physical Protection Plan AND APPUCATION TO A WESTINGHOUSE FOUR LOOP ICE CON-NUhEG-1450 AN ALTERNATIVE FORMAT FOR PATEGORY l FULL DLNSER PLANT.

CYCLE F ACluTY PHYSICAL PROTECTON PLANS NUREG/CR $745 ASSESSMENT OF ISLOCA RISK METHODOLOGY AND APPUCATON TO / COMBUSTON ENGINEERING PLANT.-

Pipe NUREG/CR 5820 AUXIUARY FEEDWATER SYSTEM R:SK BASED IN.

NUP*G/CR.4599 V01 N2 SHORT CRACKS IN PIPING AND PIPING SPECTON GUCE FOR THE STWAUNEC NUCLE AR POWER PLANT.

WELDS Semannual Report. Octoter 1990- March 1991 NUREG/7. 686$ GENERIC 6ERVICE WATER SYSTEM RISK BASED Piptng And instrumentat6on Diagram CM WE NUREG/CH $666. F AULT TREE. EVE 'If TREE AND PIPING & INSTRU~

Ouality Assurance MENT ATON DIAGRAM trEP) E DtTORS.VERSON 4 O nelerance NUREGICR-587h ASPECTS OF MONITORING AND OUAUTY ASSUR-Manual ANCE FOR HADOLABELCD ANTIE.ODIES.

Plping Wold NUREGICA 4$M V01 N2 SHORT CRACKS IN PIPING AND PIPING WELDS L.emiannual Report. Octote 1990 Marcti 1991' N 9EG/lA0072. LOFT INPUT DATASET REFERENCE DOCUMENT FOR RELAP5 VAtlDATION STUDIES Plant Transient NUREG/M 0077-RELAPS ASSE SSMENT ON DIRECT CONTACT CON.

~

NUREG4A 006a ASSFSSMENT OF THE "ONE FEEDWATER PUMP DENSATION IN HORIZONT AL COCURRENT STRATIFIED FLOW. -

P RANSIENT" IN COF RENTES NUCLE AR POWER PLANT WITF RELAPS/ MODI NUREG/lA4061: PRE. AND POST TEST ANALY$lS OF LOui MOD 2 Practice And Procettute Digest TEST $T 02 (BT 00) WITH RELAPS/ MOD 1 AND MOR2 (LOSS OF NUREGElB6 D06 R02 UNITED STATES NUCLEAR REGULATORY IEED WATE NI COMMiSSON STAFF PRACTICE AND PROCEDURE N RE I M ANALYSIS OF LOBI TEST BLO2 OHREE PERCENT a a Ju 197 June 9 COLD LEO RREAK) WITH RELAPS CODE =

Procompression NUREGAA-0037: /SSESSMENT OF RELAPS/ MOD 2. CYCLE 36 04 NUREG/CR5847. THE INFLUENCE OF PRECOMPRESSON ON THE AGAINST LOrT SMALL BREAK EXPEniMENT L3 5.

LOWTR-BOUND INITIATION TOUGHNESS OF A 533 0 REACTOR-NUREG/lA 0043 ASSESSMENT STUDY OF RELAP5/ MOD 2 CYCLE GRADE STEEL 36.04 BASED ON THE DOEL-4 MANUAL LOSS OF LOAD IEST 07 NOVEMDER 23,1985.

Prediction Model NUHEG/1A-0044. ASSESSMENT STUDY OF RELAPS/ MOD 2 CYCLE NUREGICH 500t IMPROVEMENTS IN MOTOR OPERATED GATE 36 05 BASED ON THE TlHANGE 2 REACTOR TRIP OF aANUARY VALVE DES!GN AND PREDICTON MODELS FOR NUCLE AR POWgR 11.1983.

PLANT SYS. EMS SBIR Pnaw 1 Final Oupor1Septemtw 1990-Apnl NURCO/lA'004$- ASSEESMENT OF RELAP5/ MOD 2 USING LOCE

1991, LARGE BREAK LOSS-OF COOLANT EXPERIMENT L2-5.

NUREG/lA-00a6. ASSESSMENT OF RELAP5/ MOD 2 USING SEMIS-Pressure Capacity CALE LARGE BREAK LOSS OF-COOLANT EXPE AMENT C06-3 NUREG/CH 5862. SCREENING METHODS FOR DEVELOPING INTER-NUREG4A-0049 THERMAL HYDRAUllC POST. TEST ANALYSIS OF NAL PRESSURE CAPACITIES FOR CJMPONENTS IN SYSTEMS OECO LOrT LP-FP.2 EXPERIMENT INTERF ACING WITH NUCLE AR POMR PLANT REACTOR COOL.

NUREG/lA-00$t ASSESSMENT STUDY OF RELAP6/ MOO 2 CYCtE ANT SYSTEMS.

3604 BASEb ON THE COMMISSIONING TEST REACTOR TRIP AT FULL LOAD AT THE PHILLIPSBURG 2 NUCLEAR POWER PLANT.

. Pressure Testing NUREG/lA-0058 RELAP5/ MOO 2 ANALYSIS OF LOFT EXPER6 MENT NUREGICR Sut. ExPERtMENT At RESULTS T ROM PRESSURE TEST-L9-3 ING A 16 SCALE NUCLE AR POWER PLANT CONT AINMENT.

NURE GUA-0059 ASSESSMENT OF RELAPS/ MOD 2 AGAINST NATO.

HAL CIRCULATON EXPEFUMENTS PERFORYED _ WITH THE.

Pressure vessel REWET-lit FACIUTY.

f. mEC/CR 4119 V08 N2, HFAVY-SECTON STEEL TECHNOLOGY NUREG/lA-G360. APPLICATON OF THE RELAPS/ MOO 2 CODE TO

. PROGRAM Semiarniual Progress Report f or Apnt Septemtwr 1991 THE LOFT TESTS L3 5 AND L3 6.

32 Subject Index NURE G'l ACO61 PRE. AND POST 1E S1 AN ALYS!S OF LODI MOD 2 INTE RF ACING WeTH NUCLEAR POWER P ANT REACTOR COOL-TEST ST 02 (BT@ra MTH hf L AP5tMOD1 AND MON (LOSS OF ANT SYSTEMS Ff E D W Attnj NUREWCR4870 RESULT 5 OF LWR SNUSOE R AGING RESEARCH NUni G/lA&/03 RELAPL. MOD 2 CALCULAllON Or OECD LOf T TE ST LP s W oi Reactier Safety NURE G 4A-DU64 ANALYS!S Or SE WSC ALE TEST S-LH-1 USi%

NUNEG/CP-0119 V01 PROCEE DINGS OF THE N'NETE E NTH W ATER REl AP5/ MOD 2 NE ACTOR SAF ETY INroRV AllON MEETING NUH L G / t A.0%5 ANALYb:S Or St MtSCALE 1EST S LH 2 USING N'JREG/CP40119 VD2 PROCEEDINGS OF THE NINETEENTH W ATin RF l AP5iMOD2 RE ACTOR SAFE TY INFORM ATION MEE' TING NUPE Gila ODM hELAP5/ MOD 2 ANAOSIS Or Lori E7PER! VENT NunEG/CP-0119 V03 Ph0CE EDiNGS OF THE NINETEENTH WATER lb 4 RE ACTOR SAFETY INrORMATON MEETING NURE G/lA-O%9 ASSESSMLNT OF RE LAPS / MOD 2 CYCLE 30 04 NUREG/CP 01 t9 VC3 AD PRCK4EDiNG0 N THE N:NE1E E NT H USiNG lor 1 IN16 RMEDIATE BRE AK F RPE R) MENT L$ 1 WATER RE ACTOR SAFETY IT%M ATON M TING NunL G /lA O370 ASSEb5ME N1 OF RELAPL/ MOD 2 CYCLE 30 04 WITH t OFT i ARGE BRE Ak LOCE L21 Reactor Trly NUREG/lA 0011 ANALY5tS OF THE UPTF SEPARATF El F ECTS TLST NUREGaA-00$7 ASSESSMENT STUDY OF RE LAPS / MOD 2 CYCLE 11 (5TE AM WATER COUNTERCURRENT FL OW IN THE DROkEN 36 04 BASED ON THE ^OMMISSIONING TEST RE AC10R TRIP AT LOOP HOT LE G) U$tN^s REL AP5/ MOD 2 FULL LOAD AT THE PHiLLIPSBURG 2 NUCLE AR POWER PLANT-NURE G/lA 00 74 F.E L AP5/ MOD 2 POST 1EST CALCULATION Or THE DE'CO LOrT E PE RIME NT LP.58-1 p;,,cgo, y,,,,g NUREG/iA 0076 RE LAPLe MOD 2 AN ALYSis OF A POSTi" ATED NOREG/CR M4' THE INFLUENCE OF PRECOMPRESSION ON THE

• COL D L LG $!!LOCA" SIMULT ANEOUS TO A " TOT AL DLACK OUT" LOWER-BOUND INIT!ATON TOUGHNESS OT A 531 B REACTOR-EVE NT IN THE JOSE CArinE R A NUCLE AR ST ATON GRADE STLE'L' NUREG/iA 0079 POS1 TEST ANALYSTS OF LOO! TEST BT 12 UStN3 RELAP5' HOD 2 9

?eUREG/lAumu ASSESSMENT OF RELAPS/ MOD 2 USING SE k 15-U Vit Noi-NRC REGULATORY AGENDA.Ouarlorty C INtEHMlD1 ATE tire AM LOSLOr COOLANT EsPERNENT S-thaey March 1992 NUHE G'IA-O'*2 THE ASSE S5 MENT OF RELAP5/ MOD 2 AGAINS T IVO Regulatory And Technical Report NUREG 0304 V16 Not REGULATORY AND TECHNtCAL REPORTS NLf 64 LE VANT RESULTS OBT AINED IN THE ANALYSIS (ABSTRACT INDEY JOURN AL) Annual Complaton For 19i>1 Of LOO!/ MOD 2 NATURAL CIRCULATION EyPERtMr NT A2J /A NUREG'I A 0066 ASSLSSMENT OF RLL AP5/ MOD 2 CRITICAL 6 LOW Report To Congresa MODL UE'NG MARvmEN 10 ST DAT A 15 AND 24 NUHE Gil A EA 7 RE6 AP5 MOD 2 POST' TEST CALCULATON OF THE NUnEG 0090 vid N04 REPORT '.O CON 4E$S ON ABNORMAL Of CD LO51 EFFE AlMENT LP48 2 OCCURRENCES Oclote Decemtme 1991 ret AP5/ MOD 3 Residual Heat Removal NUREG/CR 5820 CON 4EQUENCES OF THE LOSS Or THE RESIDUAL.

NUREG/CR 6820. CONSEQUENCES OF THE LOSS OF THE RESIDUAL HEAT REMOV AL SYSTEMS IN PRESSUHt2ED W ATIR REACTORS HE A' REMOVAL SYSTEMS IN PRESSURIZED WATER RE ACTORS NURE GICR-5956 THERMAL HYDRAULIC PROCESSES DURING RE.

REWET il Facility DUCED INVENTORY OPE. RATION WITH LOSS OF RESIDUAL HEAT d MEG 4A 0059. ASSESSMENT OF RELAP5/ MOD 2 AGAINST NATU-RE MOV AL, RAL DRCULAllON E y PE R! MEN T S PE RF OWAED WITH 1HE RE WE T -In F ACilli y R6nghals 4 Power Plant NUREG!lA 0038 ASSESSMENT OF TRAC PFI/ MOD 1 AGAINST AN IN.

Radiation "s pesure ADVERTENT FEEDW ATER UNE ISOLATON TRANSIENT IN THE NUHE G 0713 V11 OCCUPATONAL RADIA10N E1POSURE AT COM-4tNGHALS 4 POWER PLANT ME RCI AL NUCLE AR POWE R REAC19R$ AND OTHER F ACluTIES 19% Twen4Swond Annuat Rgort Rules NURLG 0936 Vit N01 - NRC RE GULATORY AGE NDA Qua%rly Radiation Protection ReportJanuary Match 1992.

NUME G/CR 5M9 HE ALTH PHYSICS POSITIJNS DA I A DASE Rules Of hactice N L 81 EVALUATION OF EXPOSUHE PA1Hwiv3 TO MAN FROM DI3POSAL OF RADOACTIVE 'AATERIAL; iNTO SANtTARY

' DIGEST Comrnissen. Appeal Board And Licensq Doartt 3

SM R $1 STEMS Decisons Jufy 1972 Ji w 1991 Radiolabeled Antibodies

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U EG/lA 0075 RELAP5/ MOD 2 M4ALYSIS OF A POSTULATED C

R R DO A EL D AN

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" COLD LEG SBLOC A" 5fMULT ANEOUS TO A " TOT AL BLACnOJT" EVENT IN THE JOSE CADRERA NUCLEAR ST ATION-Reactor Acender NURE G 'C6 -01 L. 31 PROCEEDING $ OF THE NtNETEENTH W ATER Safeguard RE ACTOR SAFETY INrORM AtiOh MrETING NUHEG/CP-0119 V02 PROCE EDIN iS OF THE N)NETEENTH W Al[R NUREG 1456 AN ALTERNATIVE FOR9AT FOR CATEGORY I FUEL CYCLE FACILITY PHYSICAL PROTECTON PLANS RE ACTOR S AFETY INFORM ATIO 4 MEETING NUREG/CP-0119 V03 PROCE EDAS OF THE NtNETEENTH W ATER RE ACTOR SAF ETY INFORMATO4 MEETING Safety Evaluation Report NURE GICP.0119 V03 AD PROCEEDINGS OF THE NhETEENTH NUREG,0647 509 SAFETY EVALUATION REPORT RELATED TO THE W ATER RE AC10R SAFETY IN5 04M ATON MEETING OPERATION OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND NUREGICH MQ6 APPULATON E F CONT AlNMENT AND RELE ASE 2 Doctet Nos 50 390 And 50 391(Tennesse Valley Aut%nty)

M AN AGEME NT S T R A TEG!ES TO PWR DRv-CONT AINMENT PL ANT S Safety Research NUREGICA 5875 EXPERtMENT AL MC JUNG OF HE AT MD MASS NUREG 1266 V06. NRC SAFETY RESEARCH IN SUPPORT OF REGU.

TRANSFER IN A TWO-FLU lD BUF J 4G POOL WITH APPLtCATON LATON FY 1991 TO MOLTEN CORE-CONCRETE t%TD ACTIO%

Sanitary Sewer System Reactor Conta6nment NUREG/CR 5814 EVALUATON OF EXPOSURE PATHWAYS TO MAN NUREG'CR 6732. COINE CHE WC AL FORMS IN LWR SEVERE FROM DtSPOSAL OF RADCAC9VE MATEntALS INTO SANITARY ACCOENTS Feat Report SEWER SYSTEMS Reactor Coolant system Security NUREG/CR SR62 SCREENING Mi THODS FOR DEVELOPING INTER-NW w14t6: AN,'i T ERN A* lVE F ORM AT FOR CAT EGORY l FUEL NAL PRESSURE CAPACITIES OR COMPONENTS IN SYSTEMS CYCLE F ACluTY PHISiCA PROTECTON PLANS

Subject Index 33 Setemic E ffect TLD NURE G'CP-Oii9 V01 PROCELDIN35 OF THE NINETE EN1H WATER NUREG 0637 V11 NS4 N4w TLD DIRECT RAD!ATON MONtTOh!NG RE AC10H S Ar E 1Y INF ORMATION ME F TING NETWOHK Frcpss Report Ck1*er Ewemt>er.1991 NUREG/CP 0119 VCE PROCEEDINGS OF THE NINETEENTH W ATE R NU4E G 063 7 V12 N01: NRC YLD DtHECT RADIAllON MONITORING Rf ACTOR SAF ETY INrORMArioy Mrf TING NETWORA Progssa RAut January March.1992.

NUhEG/CP-0119 V03 PROCEEDINGS Of THE NINETEE Nf H W ATER Rf ACTOR SAFETY IVORMATION MFE TING TRACBF1 NOREG/CP-0119 V03 AD Ph0CE E DINGS Of-THE N.NETEE NT H NURE G/lA-000ft ASSESSME NT OF THE 'ONE FEEDWATER PUYP WATER RE ACTOR SAFi1Y INFORMA10N MEETING ThiP TRANSIENT" IN COFRENTES NUCLE AR POWER PLANT WilH Semiscale $43 NUHEG/lA 0046 ASSE SSME NT OF RLLAPS/ MOD 2 USING bt MN TRAC PF1 CALL LARGE DREAK LOSS-OF COOL ANT E APER! MENT SM3 NUREG/lA-0050 TRACSF 1 CODE ASSESSMENT USING OECD LOF T LPJ P 1 ExPERIVE NT.

Semlocale To t NUnE G/lA 0064 ANALY5IS OF SE M: SCAL E TEST b tH 1 USING T R AC-PF 1/ MOD 1 NEL APL/ MOD 2 NURt'G/lA-0027 TRAC PF1/ MODI CALCULATIONS Or Lori ExPERi-Semiscale Test $4H-2 MENT LP 02 6 NUREG/iA 0038 ASSESSMENT OF TRAC-PF1/ MOD 1 AGA!NST AN IN-NURE G/1A 00t3 ANALYSl3 OF SEM: SCALE TIST S LH 2 LISIN3 Rf LAP 5/InOD2 AWERTENT FEE DWATER LINE 8SOLAtlON TRANSIENT AN THE

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RINGHALS 4 POWER PLANT Service Water Byatern NUREG/iA 0062 ASSESSMENT OF TRAC PF 1/ MODI AGAINST AN IN-NUhEG/CR SW; Gl NE RIC SERVICE WATE R SYSTEM RISK DAS( D ADVERTE NT PRESSURIZER SPRAY TOTAL OPEN:NG TRANSIENT INhPECTON GUIDE IN JOSE CADRERA POWER PLANT.

NUREG/lA 0018 REVIEW AND

SUMMARY

OF THAC ASSESSMENT Severe Accident

' ROM THE INTERNATIONAL CODE ASSESSMENT AND APPLtCA-NURFG 1422

SUMMARY

OF CHLHNODYL FOLLOWUP RESE ARCH TION PROGRAM OCAF )

ACTIVITif S NURE G/CR 4551 VPR1P4 E V ALUATION Or SEVERE ACCIDENT Tecton6c HISAS OUANilFICATON OF MAJOR INPUT PARAMETERS Experts =

NUREG/CR r 731: PIEDMONT SElSM!C REFLECTION STUDY. A PRO.

s Ocicewnston Of Source Tem issues GRAM INTEGRATED WITH TECTCMCS TO PRODE THE CAUSE Or NURE G/CH b732 IODINE CH( MICAL FORYS IN LWR SEVERE E AST E RN SEISMICITY ACCIDf N15 F mal Regot 4

Thermal Hydraulle Sh6pment NURE G/CR.SPSS THERMAL HYDRAULIC PROCESSES DURING RE-NUhEG 0725 Hon. PUBLIC INF ORM AllON CIA AR FOR SHtP-DUCED INVENTORY OPERA 10N WITH LOSS OF RESIDUAL HEAT MLN15 OF IHR ADI ATED RE ACTOR FUE L RE MOVAL NUREG/iA 0049 THERMAL-HYDRAULtC POST TEST ANALYSIS OF Small Break Esperiment OLCD LOFT LP FP 2 EXPERIMENT.

NUREGilA 0037. ASSESSMENT OF RELA PS/ MOD 2 CYCLE 36 04 AGAINST LOFT SMALL BREAA EXPERl MENT L3-5 Thermoluminescent Dosimeter NUREG 08.37 V11 N04 NRC TLD DIRECT RADIATON MONfTOhlNG Snubber Ag6ng NETWORK Progress Report October-December.1991 NURI G/CR 5670 F:ESULTS OF LWR SNUBRER AGING HESEARCH NUREG 0837 V12 N01: NRC TLD DIRECT RADIATION MONITORING gp,ng pgg NETWORK.Prcyees Report January March,1992.

NUREG0125 409 PUDLIC INFORMATION CIRCULAR FOR SH P-Time Step

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NL G 49 L

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TES OF COMPLIANCE FOR O" b

^

DRV SPENT FULL STORAGE casks FERENCE SCHEME FOR TRANSIENT TWO-PHASE FLOW.

Stainless Steel Tme Ust NUREG'CR 5705 EkPER! MENTAL RESULTS OF TESTS TO INVESTL NUREG 0540 V14 NO2 TITLE LIST Or DOCUMENTS MADE PUBLICLY GATE F LAW BEHAVOR OF MECHANICALLY LOADED ST AINLESS S1 EEL CL AD PLATE S AV AILABLE. February 1 29,1992 NUREG-0540 V14 NO3 TITLE LIST OF DOCUMENTS MADE PUBLICLY Standard Review Plan Av AILABLE March 1-31.1992 NUREG 1447. STANDARD HEVIEW PLAN UPDATE AND DEVELOP.

NUREG 0540 V14 ND4 TITLE LIST OF DOCUMENTS MADE PUBUCEY MENT PROGRAM IMPLEMEN11NG PROCEDURES DOCUMENT AVAILABLEApnl 1 30.1991 Steel Component Total Opening Transient NUREG/CR 5871 V01 DEVELOPMENT OF EQUIPMENT PARAMFTER NUREG/lA-0062. ASSLSSMENT OF TRAC PF1/ MODI AGAINST AN IN-ADVERTENT PRESSURI7ER SPRAY TOTAL OPENING TRANSIENT TOLER ANCES FOR THE ULTRASDN!C INSPECTION OF STE EL COVPONE NTS Appleabon To Components Up To 3 inchn Thek IN JOSE CABRERA POWER PLANT.

Stratified Flo, Training NUREG/lA 0077. RELAPS ASSESSMENT ON DIRECT CONTACT CON.

NURE G/F'R 5114 FINDINGS OF A WORKSHOP ON DEVELOPING A DENSATON IN HORLZONTAL COCURRE NT STR ATIFIED F LOW METHODOLOGY FOR EVALUATING EFFECTIVENESS Or NUCLEAR POWER PLANT TRAIN:NG Strees Corrosion Cracking NUREG/CR 5805 MET ALLURGtCAL EVALUATON OF WELD OVER, Transient Two-Phase Flow LAID PIPE SE CT ONS FROM BRUNSWICK UNIT 2 NUCLE AR NUREG!IA-00?3 TIME STEP AND MESH Si?E DEPENDENCIES IN THE POWER ST ATON HEAT CONDUCTON SOLUTION OF A SEMLIMPLICIT. FINITE DIF.

FERENCE SCHEME FOR TRANSIENT TWO PHASE FLOW Stress intensity NUREG/CR 5e60 F RACTURE MECHAMCS-B ASE D F AILURE ANALY.

Transportation SIS NUREG/CR 4831 STATE OF THE ART IN EVACUATION TIME Esil-Str e ss-Corrosion-Crac king NUREG!CR 57to STRE SS CORROS ON CRACAING STUDIES ON Transportation-incident CANDIDATE CONT AINER ALLON S FOR THE TUFF REPOSITOqY NUREG 1458. EMERGENCY RESPONSE TO A HGHWAf ACCIDENT IN SPRINGFIELD MASSACHUSETTS ON DECEMBER 16 1991.

Surface Faultin9 NUHEG'CR 5892 A HGHWAY ACCIDENT INVOLVING UNIRRADIATED NURE G>CR 4753 VOS. CANADtAN SEISMIC AGREE MENT. Annua!

NUCLEAR FUEL IN SPR!NGFIELD, MASSACHUSETTS.ON DECEM.

Report.My 1989 - June 1990 BER 16,1991.

l l

l 34 Subject Index Tuff Flepository Vendor inspection NUME G/CH4710 ST RE SSCOnRO*.,0*4 CR ACAING ST UDIE S ON NUnEGy;40 V16 Not f lCENSEE CONTRACTOH AND VENDOR IN-GANDtDATE CONTAINE R ALLOfS 6 04 THE TUFF REPOSITORY.

Spg etioN S3Atys ptpont ca3neg nep3rt,;,,v,3 y,,c n 1992 W e B W Two Fluul Dubbhng Pool NUhf G/CH 5675 ExPER MENT AL lODELING OF Hf AT AND MASS TAANSF EH IN A TWO5LUiD OUBBLING POOL WITH APPL 6CADON We6d Overtand Pipe TO MOLT (N CORE CONCHFTE INTERACTIONS NU4EG/C44M5 VETAltU4GICAL EVALUAtlON OF WELD OVER-LA!D PIPE SECTIONS FROM BHUNSWICA UNIT 2 NUCLE AA Ottrasonic inspect 60n POWER ST ATION NUHEG/CH 5871 V01 DIVELOPMENT OF EQUIPMENT PAkAMETER TOL E RANCI S F OH 1HF. ULTRASONIC INSPECTION OF SIEEL ggnp COMPONENTS Appication To Conyywnts 0910 3 Irc+t Track NURCG/CR 5114 FINDrNGS OF A WORKSHOP ON DEv[ LOPING A Uranium Mill Talling VETHODOLOGY FOR EVALUATING U f ECTivENESS OF NUCLE AR NU A[ G /C4 *,P,'A INroffMATON i 04 CONSIDERATtON IN REV!EW-POWER PLANT TRAINING ING GROUNDW ATER PROflCTION PL ANS FOR UnAN:UM MtLE T AllINGS SITES

i NRC Originating Organization Index (Staff Reports)

This index lists those NRC organizations that have publishes staff reports. The index is ar-ranged alphabetically by ma;or NRC organizations (e.g program offices) and then by sub-sections of these (e.g., divis ons, branches) where appropriate. Each entry is followed by a NUREG nurnber and title of the report (s). If further information is needed refer to the main i

Qtion by NUREG number.

AtWibORY COMMITTIE(S)

EDO OFFICE OF NUCLE AR MATERIAL SAFETY & SAFEGUARDS AGHS, ADMOHY CCNMti1(( ON RE ACTOH SAFE GUAHDS OF FtCE OF NUCt f AH M ATF HiAL SAF E f Y & S Af I GUARDS NUHL G 1327 initial DEMONS 1HAllOf4 OF THE U S 14RC'S CAPA-NUHE G 1925 V t3 A COVPIL AllON Or Hf PORTS Of THE ADVISO.

BILITY TO CONDUCT A PEHFOHMANCE ASSESSMENT F OR A HY COMMITT[E Of4 HE AC10H SAF EG'JAHDS 1991 Annual HIGH4EVEL WASTE HOOSriLYlY OFFICE OF EXECUTIVE DiktCTOR FOR OPERATIONS (EDO)

DNISON OF SAf tGUMOL & TH%itRTATON (POS1670413)

NUHLG 0725 Hoe. f\\t0LO INFCCHATION C6HCULAR f 04 SHtP-HIGIGN 1 (POST 020201)

NLHE L

f Tl F

OH CATECORY I F UEL 4i cye i t

E ur t 991 C CL E FACI I P

CA I J NUH[G 0637 V12 N01 NHC TLD DlHE CT HAD'ATON MONITORING qg NL.1WOHK Progrew Hem >r1 January-March.os92 IN SPRINGf 10LD, MASSACHUSlTTS ON DECtMBEH 16199i t

E 4 VII t I r CI ME 4T ACTIONS SIGNIFtCANT AC-70 llOtis HE SOLVE D Quadorh Pf ogen Heset JaquFyMarch 1992 ggp(3,1419 DIMEC10RY OF CEHf fFICATES OF COMPLIANCE t DO Of TICE OF ADMINISTRAtlON (PRE 870413 & POST 840's5)

DiMION 00 F HE LDOM Of itd OHMATION A PUBUCAllONS SiHV-U.S NUCLEAM RtOULATORY COMC4SION ICLS FObi fW0205 Di ntT. OF THE Gt NE HAL COUNW (POST B00701)

NUHtG 0304 V16 N04 HI gut ATORY AND Tf CHNICAL HEPCH15 N' - 403M D00 H02 UNITE D ST Air:S NUCLEAR HCGULATORY IADSlHACI INDC # JOUHN AU Annual Cornmiation F or 1WI COMMISSION STAFF PH ACTICE AND PROCEDURE NUHI G M40 V14 NO2 ilTL E llST Of DOCUMENTS MAbt PUUUC-DIGEST Commissiott Appeal Board And Licensim3 Dowd LY AVAILA01E f etwuary 1-2% IW2 Deamons July 1972. June 1991 NUHEG 0540 Vts NO3 init t LIS1 OF DOCUML NTS MADE PUBUG.

Of FICE OF THE INSPEC10R GENCHAL (POST 830417)

NUHEG 1415 V04 NO2 Of FICE Of THE INSPLCTOR L Y AV Alt AOLL Much ' St,1992 GL NE H At Semmnnual Heror10ctober 1,1991. March 31,1992 NUHEG 0540 V14 N04 Tl1LE UST OF DOCUMI NTS MADE PUDLIC.

M. NO M TAPJ D Al WAlON WN L Y AVAll ABLE Apnl 130,1g NUH( G-0544 H03 NHC COL tEb10N OF ABBRt viATIONS f4UHL G 07$0 V34 602 INOL XE S TO NUC!E AH HEGUL ATOHY COM-N[h

[kE H PHYS CS POSITONS DATA DASE M!SSION ISSUANCE S Julv Demtet 1991 NUHEG 0150 V3*i 101. INDE XE S 10 NUCLEAR HLGul ATORY COM tOO. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

M!S$10N ISSUANCES JanuNy Ma'th 1992 OF FICE OF NUCLEAH HE GUL ATORY HE SE AHCH (POST f6J720)

NUHL G 0150 V35 NO2 NUCLE AN HC GULATOHf COMM!SSON 14 NUHEG-1266 V06. NHG SAF ETY HESE AHCH IN SUPPOR1 OF "EG, SUANCES iOH FIDRUAHY 1992 Pages 4741.

Ut ATION. FY 1991.

NUHEG 0150 V35 NO3 NUCLE AR HEGUL ATORY COMM3 SON IS.

nun [G-1327 UlllAL DEMONS 1HATION OF THE U S NRC'S CAPA-SUANCE S FOR MANCH 19C2 Part 83-144 UluTY TO CONDUCT A PEHFOHMANCE ASSE SSMENT FOR A NUHLG-0750 V35 N04 NUCLLAH fdGUL ATOHY COM%5 SON IS H!GH-LEVE L WAS1E REPOSITORY DIViSON OF HEGULATORY APPtICATIONS (POST 870413)

SUANCf S f OH AF'HIL IW2 Page 145107 NUREG O713 V 11. OCCUPATIONAL HADIATION EXPOSUHL AT NUREG0936 Vit Not NHC HEGULATORY AGE NDA Charterly COMML HCIAL NUCLEAR POWER RE ACT ORS AND OTHER Heort.Janary March IW1 F ACILITIES.1961wenty-Se ond Annual Report DIVISION OF SAF E TY ISSUE ALSOLUTION (PdST 680717)

EDO Of FIC" OF THE CONTROLLf R (PRt 120410 & POST 890205)

NUHEG-1422

SUMMARY

OF CHERNOBYL FOLLOWUP HESEARCH DIVISION OF BUDGL T ACT 'VITIE S NUME G 1350 V94 NUCLE AR Fit Gut. ATOr V COMMISSON INFOH..

MATON DIGCST 1992 Eston EDO OFFICE OF NUCLE AR REACTOR REGULATION jPOST 800428g EDO OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL A

ST 7 DAIA NURL,1447. STANDARD REVIEW PLAN UPDATE AND DEVELOP.

Of FICE FOR ANALYSIS & [ VAL UATION OF OPEHAllONAL DATA. DI-ME NT PROGRAM IMPLEMENTING PROCEDURES LOCUMENT.

HiCTOR DiViStON Or HE ACTOH PROJECTS 1/11 (POST B704116 NUHf G DO90 V14 N04 HEPORT TO CONGHE SS ON ABNOHMAL NUHEG-Oh47 SO9 SAF E T Y EVALUATON REPORT HELATED TO OCCUHHLNCEh Octater Dember 1991 THE OPERATON OF WATTS BAR NUCLt A7 P'. ANT UNITS 1 AND 2 Docket Nos 50440 And 50 391 (Tennesse Ve uey Authonty)

EDO OFFICE OF INf 0RMATION RESOUf1CES MANAGEMINT & ARM DW. SON Or HEACTOR INSPE CTON & SAF EGU ARDS (POST (POST 861109) e7041t1 DNISON Or INFonMATON CUPpORT SLHViCES (POST 09020M NUMEG 0040 V16 N01 LIC[ NSE E CONTRAE (00 AND VENDOR IN-NUMEG 0910 HC2 NHC COMPRf HENSM HE COHDS OtSPOSITON SPECTON ST ATUS REPCHT Ouorterty Report. January Mwth SCHE DULE 1992 (WNtc Daok) 35

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i

NRC Originating Organization Index (Intemational Agreements)

This index lists thoso NRC organizstions that have pubbshed international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and titio of tho repor;(s). If further intornation is needed, refer to the main citation by NUREG number E DO OFFICE Of NUCL E AR REOULATOHV Rt SE A4CH (POST 5J040")

NUHLGilA-006% ANALYSIS OF SEMISfALE TESI S LH-2 US%

06 Flu OF HUCll AH HI gut ATOHY Hi SF AHC.H (POS1 Mol?O)

Hf LAPS / MOD 2 NUHEG /lA 002 / TRAC PF t/M001 cal LdLA11CNS OF LOf T E X -

NUMGliA.uu60 f tELAP5/ MOD 2 ANALYlilS OF t Of T EXPER! MENT Pl htMLNT LP 02 6 pp NUHf G/lA-0010 AN ALYSIS OF 100' IEST DLO2 (THHLE Pt HLENT NL'itLGliA 0966 ASSESSMENT Or THE "ONE f L EDWATE R PUMP COLD L EG HM AV)WITH Hi t APS CODE

/H:P TRANSIE NT" IN COf Hl.NTES NUCLE A!T POWlH PLANT NUHL G/IA O%/ ASSIS5MLNf Of HE L APS/ MOD 2.CYCLI 36 04 WH THA%f t AGAINST LOFT SMAt L HHI AK DPI HIMiN713 5 N UME GUA 00M ASSES $ME Ni Of HELAPS/ MOD 2 CYCLE 36 04 NUHF G/lA 0038 ASSESSMLNi OF THAG PF UMOD1 AGAINST AN USING LDf T INTERMEDIATE 84[ AK EAPEHtMENT iPri INADVLHil Nt fil DW AT(H LINE ISOL Ah0H T R ANSR NT IN gggggfig00/0 ASSESSME NT Of HL LAF%/ MOO 2 CYCLE 36 04 TML RINGHAL $ 4 K)W[H PLANT L2 3 WITH lOFY L ANGE OHf Ak LOCE'HE UP1F SEPARATE EFff CTS NUHEGAA UG 3 AGSISbMEN r STUDY OF HELAPS/ttOD2 CYGli NUHIG4AAWt ANAtY51S OF l 3^ O4 HASf D ON IHf (KX L-4 MANUAL LOSS Or LOAD TEST OF 1EST 1! (STE AM-WATER COUNTERCUHHENT FLOW IN 1HE HOOKEN LOOP HOT LEG) USING fif t APS/MODJ Nt i 4/

0 A [ShMENT SfilDY OF Hi L AP6/ MOD 2 CYCLE 36 0$ DASI D DN THL ilHANGb2 flEACT(M THIP Of JANUAGh 5 OH PELAf% VAUDAh0N S TUDif F I3 #8 Nt i A OD45 ASSI 5% MENT Of HELAPUMOM USiNG L(A'E Ca.WCWN 4 AM% W W W i

L AndE DHt' AK L OSS OF.COCL ANT f kPf n! MEN T L2 5 04 FEHf M EHEME rOR TR TWO PMASr FtOW NUHLG/l&Onde ASSE SSMENT Of flLL AP5/ MOD 2 USING SIMIS.

NUHEG 4A. 074 HL LAPS / MOD 2 rTEST CALCVLATION OF Calf L ARGE HREAK LliSS Of COOL ANT E XPt HtM) NT $ 0(i3 THE DE CD LOFT E XPERIM[ NT LNB 1.

NUHLGUA 0049 THLHMAL.HYDRAullC FOSI TLS1 ANAL v5lS Op Of CD t OF1 LPJ P 2 EXPtfHMENT NUHl GIL 00/5-W LAPS / MOD 2 ANALYST $ OF A POSTULATEC

" COLD !!G 58LOCA" SiMULI ANLOUS TO A " TOT AL ULACK.

NUHLG /l A-OtB0 THAC-PF 1 COOL ASSE S%MI N1 USING OEl:0 OUT" EVE NT !N 1HE JOst CABRE H A NUCt EAR ST ATION LOI T LP.F P i DPEHiMl NT.

NUHE G!iA4)057. ASSESSMLNT STUDY OF RI L AP5/ MOD 2 CYCt E NUHE G/tM)077 HELAPS ASSE SSML NT ON OlHECT CONTACT

.g 3604 DASLD ON THE COMMISSIONING TEST HEACTOH THP Al CONDENSATIOtt IN HOHl/ON1 AL CUCURHENT STHATIFIED I ULL TOAD AT THf PHILLIPSUUHG 2 NUCLE AF 'OWI.R PLAN 1 f I OW-NUHL GAA Othe Rf L APS/ MOD 2 ANAL YSIS Of wOf Y DPEH%IlNT NUHEGUA 0078 HEVIEW AND SUMMAk' OF THAC AS$f SSMENT IHOM DIE INTERNA 1 TONAL CODE ASSLMMENI AND APPLICA.

t [t 3 NUHt G4A.0059 ASSESSMf NT OF HE LF5/M' 4 AGAINST NATU-TlON PH(XiRAM (CAP)

HAL CIRCUL AllON EXPERIMENTS I cRI JDMED WITH THE NUMEG4A-007tt POS T.T EST ANALYSIS Of LOOL TCST DT 12 HIWET-lit f ACluf Y.

USING HELAPS/ MOD 2 NUHI.G4A 00tO APPLICATION OF THE HELAP5./ MOD 2 CODE f 0 NUHLGAA 00M ASSESSMEN1 OF HELAPS/ MOO 2 USING SEMIS-THE LOf f 11 STS L3 5 AND f fl 0 CALE INTEHMtDIATE DALM. LOSS 00 COOLANT EXPERIMLNT NUHE G 'IA 0001 PRL. AND POSI list ANALYSIS OF LOGI MOD 2 Sin 3 TE ST ST 02 (BT.00) W *H HELAP5/ MOD 1 AND MOO 2 ttOSS OF

.1UHtGUA 0062. THE ASSESSMINT OF REl.APS/ MOD 2 AGAINSI F EI D W ATl H)

IVO L OOP -SE AL TESTS NUHLGilA,0062 ASSESSMENT OF THAC Pf i/ MOD; AGA!NST M NUREGnA-nr$4 HitEVAtJT HESULTS OOTAINED IN THE ANALY.

lHADVfEllN1 PHFSSUHl2LR SPHAY TOTAL OPLNING TRAN-

$15 Or t.Ohl/ MOO 2 NATURAL C'HGULATION DPERIMf NT A1-SIFNT IN JOSE CADHEHA POWEH PLANT 7/A NUHLG/lA 0063 HE L APUMOD2 UALCUL AKN Of OECD LOCT NUHEGNAf006 ASSESSMENT OF HLLAPS/ MOO 2 CfUtiCAL FLOW TLST L PJW 01-

• 40 DEL USING MAHVIKEN TEST DATA 'S AND 24 NvHLG/lA t M ANALYSIS Or SEMISCALE TE ST Sl H 1 USWG NUB (GNA0047 HELAPS/ MOD 2 POST. TEST CALCLE^ TION OF RLt. AP5 A ;',1

'HL OCCD LOh DPERIMENT LP-SD 2.

37

NRC Contract Sponsor Index (Contractor Reporto)

This index lists the NRC organizations that sponsored the contractor reports listed in this compilation. It is arranged alphabetically by major NRC organization (e.g., program offi,ce) and then by subsections of these (e.g., divisions) where appropriate The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) preaared by that organi-zation. If further information is needed, refer to the main citation by the VUREG/CR number.

M0, OFFICE FOR ANALYSIS & EVALUATION OT OPERATIONAL NUREG/CR 5797-IVPACTS DRC. VERSION 21 Codes And Data Ven.

DATA hcation.

OFFICE FOR ANALYSIS & EVALUAllON OF OPERATONAL DATA, DI-NUREG/CR5814-EVALUATION OF EXPOSURE PATHWAYS TO RE CTOR MAN FROM DISPOSAL OF RADIOACilvE MATERIALS INTO SANb

)

NUREG/CR 2000 VII N2 LICENSEE EVENT REPORT (LER)

TARY SEWER SYSTEMS l

COMP!L AT60N For Month Of f ebruary 1992 NUREG/CR5077: ASPECTS OF MONITORING AND OUAUTY AS-1 NUREGICR 2000 V11 N3 LICENSEE EVENT REH3RT (LER)

SURANCE FOR RADIOLADELED ANTIBODIES i

COMPILATON For Month Of Merch 1992 OtVISION OF SAFETY GSUE RESOLUTION (POST M0717)

EDO. OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS R SKS U T1 AT N M

I P DIVISON OF SAFEGUARDS & TRANSPORTATION qPOST 870413)

NUREG'CR 6602-A HIGHWAY ACCIDENT INVOLVING UNiRRADIAT.

PARAWRS Egeds' hmnation Of Source im !ssues NUREG/CR5301: MODEL S AND RESULTS DATADASE (MAR-ED NUCLE AR FUEL IN SPRINGFELO.MASSACMUSFTTS ON DE-D) VERSION 4 C Geference Manual CEMBER 16 1991 DIVISION Of LO'W LEVEL WASTi MANAGEMENT & DECOMMISSION.

HUREG/CR-5004 V01: ASSESSMENT OF ISLOCA RISK METHOD-,

OLOGY AND APPLICATIUN TO A BADCOCK AND WILCOx NUCLE.

ING (POST 870413)

NUREG/CR 5858. INFCRMATION EOR CONSIDERATION IN RE.

AR POWER PLANT. Man Report VIEWING GROUNDWATER PROTECDON PLANS FOI-URANtUM NUREGICR5604 V02 ASEiSSMENT OF iSLOCA RISK - METHOD-MILL TAluNGS SITES.

OLOGY AND APPLICATION TO A SARCGCK AND WILCOx NUCLE-AR POWEA PLANT. Append:ces A H EDO + OFFICE OF NUCLEAR REGULATORY RESEARCH (PCC 820405)

NUREGICRS604 i 11 ASSESSMENT OF ISLOCA RISK - METHOO-DIVISION OF ENGINEERING (POSI 8m41M OLOGY AND APPi tC ATION TO A BABCOCK AND WILCOX NUCLE NUREU/CR4219 VD8 N2: HEAVY 4ECTION $1 EEL TE OtCGY AR K7WER PLANT.4ppendices i M.

PROGRAM Sennannual Propress Report For Apnl Setten.ser 1991 NUF{G/CR4744 ASSESSMENT OF ISLOCA RISK-METHOUOLOGY NUREG/CR4469 Vt2. NOWSTRUCTIVE EXAMINATION (NDEI RE-AND APPLICATION TO A WESTfNGHOUU TOUR LOOP ICE CON-LIADILITY FOR INSERVICE INSPECTON OF LIGHT WATER DENSFR PLANT.

REACTORS Annual Repn:1. October 1989 September 1990 NUREG/CR5745= ASSESSMENT OF ISLOCA RISK METHODOLOGY NUREG/CR4539 V0f N2' SHORT CHACKS IN PIPING AND PANG AND APPUCATION TO A COMBUSTION ENGINEERING PLANT, WE LDS Semannual Repmt October 1990 March 1991.

NUREG/CR6866. FAULT TREFEVENT TREE.AND PIPING & f4 NUREG/CR4753 V05-CANADIAN SEISMIC AGREEWNT. Annual STRUMENTATION DIAGRAM (FEP) EDITORS, VERSION 4 0 Refer.

Report. July 1989 + June 1990.

NUREG/CR 5t21 EXPERIMENTAL RESULTS FROM PRESSURE OfVI N F YSTEMS FtESEARCH (POST 880717) l-TESDNG A 16-SCAL E NUCLEAR POWER PLANT CONTAINUENT-NUREGICR 5114 FINDINGS OF A WORKSHOP ON DEVELOP 1N3 A NUREG/CR4445 PERFORMANCE OF INTACT AND PARTIALLY DE-METHOOOLCGY FOR EVALUATING EFFECTIVENESS OF NUCLD GRADEO CONCRETE BARRIERS IN UM! TING M ASS TRANSPORT.

NUREG/CR5693 AGING ASSESSMENT OF COMPONENT COOUNG NURE 573 NE HEMICAL FORMS IN LWR SEVERE W AT ER SYSIE MS IN PRESSURIZED WATER REACTORS NUREU/CR5720: MOTOR 4PERATED VALVE RESEARCH UPDATE ACCIDENTS. Final Report NUREG/CR 5805 IDENTIFICATON AND ASSESSMENT OF CON-NUREG/CR5 ?31 PIEDMONT SEGMIC REFLECTION STUDY; A PROGRAM INTEGRATED WITH TECTONICS TO PROSE THE TAINMENT AND RELEASE MANAGEMENT STRATEGIES FOR A CAUSE Of E# STERN SE.SMICITY.

RWR MAMK 11 CONTAINMENT, NUREG/CR 5785. EXFtRIMENTAL RESULTS OF TESTS TO IN ES.

NUREG/CR 5806 APPUCATION OF CONTAINMENT AND RELEASE TIGATE FLAW BEHAVOR OF MECHAN!CALLY LOADED 81. %

MANAGEMENT STRATFGIES TO PWR ORYMONTA;NMENr LESS STEEL CL AD PLATES PLANTS NUREG/CR580h IMPROVEMENTS IN MOTOR OPERATED GATE NUREG/CR 5855: THERMAL-HYDRAUUC PROCESSES DURING RE-VALVE DESIGN AND PREDICTION MOEWLS FOR NUCLEAR DUCED INVENTORY OPERATION WITH LOSS OF RESIDUAL POWER PLANT SYSTEMS SalR Phase i Final ReportSeptember HEAT REMOVAL.

1990 Apnl Iggt MUREG/CR5875. EXPERIMENT AL MODEUNO OF HEAT ANC MAS $

NURE G/CR 5647. THE INFLUENCE OF PREC(,wtPRESSON ON THE TRANSFER IN A TWO FLOO OUBBUNG POO'. WITH APPLICA-LOWER 00UND INITIATION TOUGHNESS OF A SW B REACTOR.

TON TO MOLTEN CORE CONCRETE INTERACTIONS.

GRADE STEEL NUPEGICR$MO FRACTURE-MECHANICS-BAS 'D FAP URE ANAL.

EDO - OFFICE OF NUCLEAR REACTOR REGut ATION (POST 800428)

YSis OlvlSION OF ENGINEE8ING 'ECHNOLOGl (PO!1690027)

NURLG/CR$362-SCREENING METHODS FOR DEVELOPING IN NUREG/CR 5805. METALLURGICAL EVALUAllON OF WELD OVER-TERf Al PRESSURE CAPACITIES FOR COMPONENTS IN SYS.

LAID PIPE SECTONS FROM BRUNSWICK UNIT 2 NUCLEAA TEMS INTERFACING WITH NUCLEAR POWER PLANT REACTOR POWER STAT'ON.

COOLANT SYSTEMS DIVISION OF SYSTEMS TECHNOLOGY (POST 800027)

NUREG/C4587J RESULTS OF LWR SNUBBER AGING RE$EARCH NUREG/CR 5820- CONSEQUENCES OF THE LOSS OF THE RESIDi NUREG/CR 5671 VOI: DEVELOPMENT OF EOU!PMENT PARAME.

LAL HEAT REMOVAL SYSTEMS IN PRESSURt2ED WATER REAC-TER TOLERANCES FOR THE ULTRASONIC INSPECTION OF TORS STEEL COMPONENTS Apphcation To Componunts Up To 3 inches DIVOlON OF RAOtATON PROTECTION & EMERGENCY CREPARED.

The NESS POST 670411)

DIVISION OF REGULATORY APPUCATIONS (POST 8704t3)

NU9EGICR4831. STATE OF THE AR T IN EVACUATON TIME EST!.

NUREG/CR 5672 V02, CHARAClERISTICS OF LOW LEVEL RADIO-MATE STUO!ES FOR NUCLE AR POWEA PLANTS ACTIVE WASTE Decontannnaton Waste Program Annual Report NUREG/CR 5569 HEALTH PHYSICS POSITIONS DATA BASE.

Foo Frscal Year 1991 NUREG /CR4633 AUX!UAR f FEEDWATER SYSTEM RISK. BASED NUREG/CR 4710: STRESS-CORROSION-CRACKING STUCiES ON INSPECTION GUIDE FOR THE TURKEY POINT NUCLEA9 POWER CAND:DATE CONTAINER ALLOYS FOR THE TUFF REPOSITORY PL AN T.

39

40 NRC Contract Sponsor inder NUAE O 'CR '#M t : AUXILIARY F E LDW A.iLR Sf $f f M H:SK BASED NUGE G/CH 56% GE N.itC SERVCE W ATE H SY ST E M HISO INSPwCilON GUIDE FOH TH11 P.[WAUN! L NUCLEAR POWER B AM O INSP( CTON GUIDE 8'l ANT.

i b

d i

i

Contractor index This index lists, in alphabetical order, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and titles of their reports. If further information is needed, refer to the main citation by the NUREG/CR number.

ARIZONA ST ATE UNIV, TEMPE, AZ CORTEST COLUMBUS TECHNOLOGIES, INC. (FORMERLY CORTEST NUREG/CR 4551 V2R1P4 EV4UATION OF SEVERE ACCIDENT COLUuDUS,1NC,)

RISKS OUANTtF6 CATION OF M AJOR INPUT PARAMETERS Euperts' NUREG/CR 57t0 STRESS CORROSION-CRACKING STUDIES ON Determrshon Of Source Term issues CANDtDATE CONTAINER ALLOYS FOR THE TUFF REPOSITORY, BATThlE MEMORIAL INSTITUTE DDL OMNI ENGINEER!NG CORP.

NUREGICR 5660 FRACTURE &ECHANICS-BASCO F AILURE ANALV-NUREG/CR 5114 FINDINGS OF A WORKSHOP ON DEVELOPING A Sf3 METHODOLOGY FOR EVALUATING EFFECTIVENESS OF NUCLEAR BATTELLE MEMORIAL INSTITUTE. COLUMBUS t ADOR ATORIES NUREG/GR-4599 V01 N2 SHORT CRACKS IN PIPING AND P PING EG&G IDAHO, INC.

WELDS Semiannua Report October 1990 March 1991-NUREGICR-5301' MODELL AND RESULTS DATABASE (MAR-BATTELLE MEMORIAL INSTITUTE, PACIFIC NORTHWEST NU M5 PE FO OF INT ACT AND PARTIALLY DE-LABORATORY GRADED CONCRETE BARRIERS IN LIM 6tNG MASS TRANSPORT.

NLREG dCR-4169 V12. NONDESTRUCTIVE EX AM; NATION (NDE) REll-NUREG/CR-5604 V01. ASSESSMENT OF ISLOCA R'SK METHODOL-A8illTY FOR INSERVICE INSPECTION OF (,HT W ATE R OGY AND APPLICATION TO A BADCOCK AND WILCOX NUCLEAR HE ACTORS Annuai Report Octaber 1989 September 1990 NUHEG/CR 4831 STATE OF THE ART IN EVACUATION lime ESTi~

NU EGiCR 5 4 0 SMENT OF ISLOCA RISK METHODOL-OGY AND APPUCATION TO A BABCOCK ANO WILCOX NUCLEAR N E R5 3 11 Y FEEDW TR YSTEM RISK. BASED IN.

I E L^

^ P*

$^

EPE ION GUIDE FOR THE TURKEY POINT NUCLEAR POWER 5

SSMENT OF ISLOCA RISK METHODOL-OGY AND APPLICATION TO A BABCOCK AND WILCOX NUCLEAR NUREG/CR 5614 EVALUATION OF EXPOSURE PATHW AYS TO MAN POWER PLANT Append 4ces t M.

FROM P*POSAL OF RADIGACTid MATERIALS INTO SANITARY NUREGiCR 5672 V02 CHARACTERISTICS OF LOW LEVEL RADIOAw, SEWER SYSTEMS NUREG/CR 5821 AUXiUARY SEEDWATER SYSTEM RISKMASED IN.

TlVE WASTE Decontamination Waste Program. Annual Report For Fmcai Year 091.

SPEC ~lON GU!DE FOR THE NEWAUNEC NUCLEAR POWER PLANT.

N'JREG /CR 5720: MOTOR OPER ATED VALVE,,ESEARCH UPDATE NUREGl.;R-Sa58 tNFORMATION FOR CONSIDERATION IN MEVIEW.

NUREG/CR-5744. ASSESSMENT OF iSLOCA RISK METHODOLOGY ING GROUNDWATER PROTECTION PLANS FOR URANIUM MILL AND APPLICATION TO A WESTINGHOUSE FOUR-LOOP ICE CON-TAlltNGS SITES NUREG/CR-5870. RESULTS OF LWR SNUBBFR AGING RESEARCH DENSER PLANT NUREG/CR 5671 V01: DESELOPMENT OF EQUIPMENT PARAMETER NUREG!CR-5?S. ASSE%SMENT OF ISLOCA RISK-METHOOOLOGY TOLERANCES FOR TME ULTRASONIC INSPECTION OF STEEL AND APPLICATION TO A COMBUSTION ENGlNEERING PLANT.

COMPONENTS Arp4catan To Components Up To 3 inches Th.ck NUREG/CR 5820 CONSEQUENCES OF THE LOSS OF THE RESIDUAL HEAT REMOVAL SYSTEMS IN PRESSURl2ED WATER REACTORS BROOKH AVEN N ATION AL LABOR ATORY NUREG/CR-5855 THERMAL HYDRAULIC PROCESSES DUR!NG RE-NUREG'CP4119 VOI PROCEEDINGS OF THE N!NETEENTH WATER DUCED INVENTORY OPERATION WITH LOSS OF RESIDUAL HEAT REACTOR SAFETY INFORVATION MEETING REMOVAL NUREG/CP-0119 V0E PROCEEDINGS OF THE NtNETEENTH WATER NURE3/CR 5862: SCREENING METHODS FOR DEVELOPING INTER.

REACTOR SAFETY INFORMATION MEETING N AL PRESSURE CAPACITIES FOR COMPONENTS IN SYSTEMS NUREG/CP 0119 V03. PROCEEDINGS OF THE N!NETEENTH WATER INTERFACING WITH NUCLEAR POWER PL,MT REACTOR COOL.

REAC1DR SAFETY 'NFORMATION MEETING ANT SYSTEMS NUREGICP 0113 V03 AD PROCEEDINGS OF THE N! NET E ENTH NUREG/CR 58E5 GENERtc SERVICE WATER SYSTEM RISK-BASED W ATER RE ACTOR SAFETY INCORMATION MEETING INSPECTION GUIDE NUCEG/CR 5693 AGING ASSESSMENT OF COMPONENT COOUNG NUREG/CR-5866. F AULT TREE _ EVENT TREE,AND PlPING & INSTRU-W ATER SVSTEMS IN PRESSURI2ED WATER REACTORS MENTATION DIAGRAM (FEP) E DITORS. VERSION 4.0 Reference NUREG/CR 5805 ICENTIFICATION AND ASSESSMENT OF CONTAIN-Ma%al MENT AND RELEASE MANAGEMENT STRATEGIES FOR A BWR EOE ENGNEMG CONWAMS WMEM EOE EMMM NU E R 86 AT ON OF CONTAlNMENT AND RELEASE M

GEMENT STRATEG!ES TO PWR DRY CONTAINMENT N E /CR 5867 SCREEN!NG METHODS FOR DEVELOPING INTER-NAL PRESSURE CAPACITIES FOR COMPONENTS IN SYSTEMS NUREG/CR 6875 EXPERIMENTAL MODEUNG CF HEAT AND MASS INTERFACING WITH NUCLEAR POWER FLANT REACTOR COOL-TRANSFER IN A TWO-FLU!D BUBBLING POOL WITH APPLICATION ANT SYSTEMS.

TO MOLTEN CORE-CONCRETE INTER ACTIONS NUREG/CR-5877 ASPECTS OF MONWOR:NG AND OVAL" ' ASSlJR-GR AM. INC, ANCE FOR RADIOLABELED ANTIBOD4ES NUREG/CR 5797 IMPACTS-BRC. VERSION 2.1. Codes And Data Venfi-NUREG/C458P5 METALLURGICAL EVALUATION OF WELD OVER LA!D PIPE SECTIONS FROM BRUNSW CK UNIT 2 NUCLEAp catK>rt KALSI ENGINEERING, INC.

CANADA NUREG/CR 5607. IMPROVEMENTS IN MOTOR OPERATED GATE NUREG/CR 4?$3 V05: CANADIAN SEISM C AG R EEVENT. Annug VAL,E D7GN AND PRED!CTION MODELS FOR NUCLEAR POWER PLANT SuTEMS SB R Phase i Final Report September 1990 Apnl ReportJuly 1C60 June 1990 1991 CAN ADIAN COMMERCIAL CORP, NURFG/CR4753 VOS CAN ADI AN SEISMC AGREEMENT Annuat LAKE ENGlHEERING CO.

Repc4 Ju y 1900 - June 1990 NUREG/CR 5E70 RESULTS OF LWR SNUBBER AG;NG RESEARCH.

r 41 l

d 42 Contractor index LAWHENLE LivERMORE NATIONAL LADORATORY NUREG/CR 5860 FRACTURE MECHANICS-BASED FAILURE ANALY NUREG/CR %92 A HIGHWAY ACCIDENT INv0LVING UNIHRADIATED Sls NUCLE AR F ULL IN SPRINGf lELD. MASSACHUSETTS,0N DECEM SAFETY & RELIADILITY OPTIMlZATION SERVICES,INC.

Of R 16,1991 NUREG/CR 4551 V2A1P4 EVALUATION OF SEVERE ACCIDENT RISKS QUANTIFICATION OF HAJOR INPUT PARAMETERS Egerts' MARYLAND, UNIV. OF, COLLEGE PARK, MD NUREG'Ch 5847 THE INFLUENCE OF PRECOMPRESSION ON THE Deterfrunation Of Source Term lasues t WER0 ND If CTIATION TOUGHNESS Or A 533 0 REACTOR' SANDIA NATION AL LABORATORIES NUREG/CR 4551 V2 RIP 4 EVAL.UATION OF SEVERE ACCIDENT RISKS OUANinflCA !ON OF MAJOR INPUT PARAMETERS Enerti MAZOUR ASSOCIATES,INC.

NUREG/CR 5114 FINDINGS OF A WORKSHOP ON DEVELOPING A NUF

(

l TL ESULTS FROfA PAESSURE TEST.

ING A itSCALE NUCLEAR FOWER Pt. ANT CONTAINMENT.

ER

' TRA G

• GALGtCR4797. IMPACTS DRC,VLRSION 21 Cooms And Data Venh-MiHNESOTA, UNIV. OF, WINNEAPOLIS MN NUREGrCR M77. ASPf CTS OF MONITORING AND QUAllTY ASSUR-SCIENCE APPLICATIONS INTERNATIONAL CORP,(FORMERLY ANCE FOR RADIOLABELED ANTIBODtES.

SCIENCE APPLICATIONS, NUREG-0713 V10 OCCUPATIONAL RADIATION EXPOSURE AT COM-OAK RIDGE NATIONAL LADORATORY ME RCIAL NUCLEAR POWER REACTORS AND OTHER NURE GICR-2000 Vit N? LICE NSEE EVENT REPORT (LER)

F ACILITtE 5.1089 Twenty-Second Annual P ' port COMPILATION For Month Of February 1092.

NUREG/CR 4551 V2R1P4 EVALUATION OF SEVERE ACCIDENT NUREG/CR 2000 V11 N1 LICENSEE EVENT REPORT (LER)

P'SKS OUANTIFICATION OF MAJOR INPUT PAHAMETERS Experts' COMPILAflON f or Month Of March 1992.

Determination Of Source Term isves NUREG/CH.4219 V08 N2 HEAVY-SECTION STEEL TECHNOLOGY PlOGRAM Semiannual Progress Report For Apr4SoMenita 1991 TASK GROUP ON DIGIT ZATION OF INDUSTRIAL RADIOGRAPHS NJ*H G/CR 55G HE ALTH PHYSICS POSITIONS DAT A DASE.

NUREG 1452 REVIEW AND EVALUATION OF NUREG,CR 5732 IODINE CHEMICAL FORMS IN LWR SEVERE TECHNOLOGY,EOUIPMENT, CODES AND STANDARDS FOR DIG!Ti.

ACCIDENT S Final Roport.

ZATION QF INDUSTRIAL RADtOGRAF' HIC FILM NUREG/CR 5785: EXPERIMENTAL RE SULTS OF TEETS TO INVESil-GATE. FLAW BEHAVIOR OF MECHANICAtt.Y LOADED STAINLESS VIRGINIA POLYTECHNIC INSTITUTE & ST ATE UNIV., BLACKSBURG, STEEL CL AD Pt.ATES VA NOREG/CR&47 THE INFLUENCE OF PRECOMPRESSION ON THE NUREG/CR 573

• lEDMONT SE'SMIC REFLICTION STUDY; A PRO-LOWER BOUND INIT6ATION TOUGHNESS Of A 633 D REACTOR-GRAM INTEGR, TED WITH TECTONICS TO PROBE THE CAUSE OF GRADE STEEL E ASTERN SEIS # CITY L__.

international Organization Index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation. Listed belo,w each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number, DL LGIUM T h/ Cli[q NUHLG/1A-tm/ L! L PfqMOD2 PO".T-TE"1 CMCULATIOh NUHrG1A 043 /6'.f t W NT S TUDY Of PE L APMMOD2 M 1HE Of CD LOi T LXF I RiMLNT LP CB{

CYCt i % N l!Al'i D ON IHf UO{ l ~4 IMUL LOS; OF HIDROf CT!"CA ESP' NOi A LOAD TIST Of NOVLMUf a p NURLG4A W TM!NT OF THE ONE FLtDWATER fiUle G1A LO44 AMEnef NI. MD ( Of FE L apt < MOD 2 PU'J P TDW THANDE NT" IN CO!RENTES NUCLE AR CYCL [ 30 05 0AM D UN iH! NHAML J hl Af"Ok ins POWLR F LANI YnTH 1hACT@ l Of JAf rJ Ah r 11.19 i TE CNATOM. S A NUHEGMA ON AN W NT Of lfsAC Pf t ' MOD 1 iE DL RAL Rt PUBLIC Of Gt HMANY AGAC AN MN W M N WH G WW W AL WON [t KK A ll NZ.NI IN JfE{ C AN;! H A POWi h PLANT.

OflNdU EA C E SiE M! tri AG 1 I f d iCH i NE hhl t M Ut iUNb m./J; A ll A NUREGIA M U ASLt<MLNi ' T UO f Of hE L A PD M OD2 CN Ci l 36 04 DA!! D ON T Hi COMM DON:NG TFST tiunrG1AS JD 1 E L APf ' MON AN ALY: '. OF A TOSTU-hf ACIOP 1R:P AT f W i ( D A ' A i llR I WilPCDUR,,

LAllD MD t m WWN WN.WW m A 10-NUCLi A4 POWi H PLAN; TAL N A3 O G' Wi NT IN h R C CAR RA NW CLE AR SI AllON flNLAN D IMATRAN VOlMA OY OVO; SWIDEN NUHF GaA (CEO 1 Hi A:Mt ' i N I u ' E LAP! MOD; SWI DDI NUCL[ AH POWL R INSPL CIChAlL AGAltCT IVO L OOP OL At li ST NURE G/l A ' CE ACTEFME NT Of RELAP5/MGD2 CYCLE T E CHNIC AL Nf L ARCH Cf nth ol fiNL AY) 36 04 AGAINST LOF T EMALL [mErA E xi fiqMENT t 3 5 NURE ChA- (oui AZ! ZM! fit OF f(L AIT MT2 / h :NOT NUREGSA-C E A"SiEEMENT OF T RAC -Pr 1/ MOD 1 N AT Uh Al C!RGUL AllON Df L HME NI5 V E DFORMLD AGAtNGT ANIMADVi HiLNT F f LDWATi RLINEICOLAllON W;iH TH[ h!%E T-la F AC!UTy TRANSIENT IN THE RiNGHA:% 4 POWER PLAoT ITALY UNIVE 8011 A DLGti SIUni Di mA UNiiED KINGDOM NUREG1A 05) RLLl% 4NT hl!MT" On1MNED IN TIK Cr NiRAL t LE CTfaCfrY GENWilNG tiOAhD ANAL YCG Of (OM MOD : NAlURAL G CUL AI60N EL NUHt G1A-00% AN ALYE!SOF L OU:1 L ST Ut 02 (I Hai E PL R-PI rim [ M A/, / A CENI COLD LEG OhE AN W;TH HEi. Art CODE trJRE G'l A M 1 PRE. AND f OCT.TFLT ANALYStG OF LOGI REPtlBLIC OF CHINA MOC2 TEST ST 0222 (DT un WMU M mt.PL MODI AND INSilTUTE OF NUClf AR I Nr RGY RL. LCH MOP 2 ROU Of f LI D WATL Ri NUPf G1A N15 A$i rim! NT C1 hFlF M: 0-U94G N ATIONA! PCMH

~

1 OCl LA RGl life AK 1 LM i DF COOL ANI L a Pt hNi NT NUREG5A@%Q APPtICATION Of THL hEl APE' MOO 2 CODE 12 5 TO lHE LOFT TLSE t M AND tM fiUNE MA-M46 ANLi W NT Of RE LAP. MON UK NUME G.q A DE RE L APWMOV LAICULAllON Of OE CD St MdCAU L Ak i iMAtLOG Of 4 ON ANT EFUU LOFT TE ST LP IW 01 MENT SN 3 Nt,U G/1HR4 ANALY:1 W-S{ MR/M 1GT bl H-1 US NW G1 A - E3 AEnE? M N1 Or hEl/t.M Q:0 UC M ING RELAIGMOD2 EL MMCALL INil HD AT f Id k it if GOO!.ANI NURE31A -Nf5 AN ALYSM Of El MSCAL! IEST SLH-2 US-

[ XP1 R:Mt NI L !? '

lNG pty ruOo2 NUREG1A-CM Ril APNMUD2 ANALYC8 01 L OFT EXPE RF REPUBLIC Of KOHE A ME NT [9 4 WORL A IN TilUTE OF Nt CL L AR EM L Ty WINF RITH TECHNOLOG ( CENTRE NUREG;1A %69 ATEOWiNICF Hil. Af9 MOD 2 CYCLE NURE GMA-Cf2/ T HAC-f f 1/ MOD 1 CALCUL AllON^i CF LDFT 36 04 UNN3 L OF T inh HMEC! ale EHLA% E XPLH:w NT EXPER: MENT LP {2 0 t%1 NURE GAA-00L8. RELAPL/MGC2 ANAL"E Of 1 OFT F XPL Ri-NUREG4A4WD ASSEL 3 MENT OF HELAP5 MOD 2 CYCLE Ml 4 f lD-3 M 64 VATH LOT T L ARGE ImL Ah l OCE L2 ^

NUREGliA lW1' AN?J YS S Ct THE UPlf CEPARATE EF, NUREG;1A W hi LAPS AEW MENT ON DSE CT-CON-F ECTS TEST (STE Au-W ATE R COUNTE RCUhRE NT FLOW T ACT CONDLN".ATION IN HnN tNTA< COCURRE NT IN Tdf EROKEN LOOP HOT 1 EG; USING NF t APEJMOD2 9TRAlli If D FLOW NUREG/1A-0072 LOFT INPUT DATASL T PLTERENCE DOCU-NUR!G/lA-GM ADSLSSMI NT OT REl AP5M?.D2 CRRICAL ME NT F OR RELAPS VAUDATION STUD.EC FLOW M,)DEt U3NG MMMhEN TEST DAT A 15 AND 24 NUREGAA-EW3 llML STEP AND UESH Si/E DEFENDEN-CES IN THE HE AT CONDUCTION SOLUTION OF A SEMI-IMPUCIT, F INtTE DirFERENCE SCHEME FOR TRANQENT S P AIN TWO PHACE FLOW CL NTRO DE INVEST;GACIONES [ MLDGETICAg UURECAA~CC9 POST 'IECT ANALYL3 OF LOO! TEGT BT -12 NUREGMA 4E0.RAC-PF1 CODE ASSEESMENI U0iNG USN3 HELAPLMOD2 OECD 1Of T LP F P-1 EVOLRWENT CONSFJO DE SEGUFUDAD NUCLL AR NUDEG1A-OJ47 THERMAL-HYDRAUUC POST-TEST AN ALY-UNITED ST ATES SIS OF OECD t Of T LP-F P 2 i APL RIME NT LOS ALAMO 9 NATIONAL L; ORATORY NURLGaA-00/4 HELAP5 MOO 2 POST-TEST CALCULA1 TON NURI GqAM8 RIVIEW nND

SUMMARY

OF TRAC ACSESS Of lHL DiCD LOf T EXPEfCM!NT LP 00 -1 MENT f'HOM THE INTEfiN ADON AL COCE /CSESSL ENT AND APPUCATiON PPGGhAM OCAP; 43

_ _. _ = _ _ _ _ _ _ _ - _ _ _ _ - - - _ _ _ _ - _ - _ - - - _ _

I 4

4

_ - - ~ ~

---

~~' -

Licensed Facility Index This index lists the facilities that were the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical order. They are preceded by their Docket number and followed by the report number. If further information is nooded, refer to the main citation

.by the NUREG number.

Br ened Yom Ekint 8%at UrW 7 CwuMa h@EGICR f tes

% 251 Turkey hwit hart Urst 4. Fudit F%* arx!

NGHLG'CR %33 50 324 v

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% Ya newerve Naive Pt*t' ha<t Wyonw PM f(JHCG/ Lit *,821 50 110 Wa% BW Nutiear ha1. U d 1, Ternsp4 NUREGWT Set)

SerK8C#0 Viikiy AuthCr'ty 50 2')O Iwfh4W kUNil hdMt. l.h! 1 f kAd8 P Nrsif 5fiff NVIN G, CR 6C33 60 391 W3'14 Baf N3MW Platt Uni $, fe8Vnwe NUfM084? 509 C

tyd Co Va!ity A.ithatty i

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1

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0.$ NUCLE AR REGULATCJ4Y COMMISSION

1. REPORT NUMBE84

( Assegmd f)y NHC Add Vol.

G49)

Nr4CM 1102, Supp., dev., and Addendum Num.

32m. m BIBLIOGRAPHIC DATA SHEET

• • * < H *"Y I as e niruci e sont>. revwsei NUREG-0304 Vol.17, No. 2

2. m Lt ANu t,ve m tt 3 DAll HLPOHI PUtttlSHLD Regulatory and Technical Reports (Abstract Index Journal)

MON T H YEAR Compilation for September 1992 Second Quarter 1992 4 nN oH c.HANT NuMus April-June 6 AVIHUHtbj 6 lYPtOFHEPOHr Reference

7. Penico covEnto oncwme oawu April-June 1992 11 Pt HP OHMsNO OHG Ar4t2 A T ION - NAML AND AC-)HL SS Of fetC, provvJe DivitKm. Office or Hegron, U. S. Nuctedr Haagalatory Comrnession, ard maileng address if contractor, provuk narne and mathng address }

Division of Freedom of Information and Publications Semces Office of Adminiatration U.S. Nuclear Regulatory Commission Washmgton, DC 20555 9 66*ON'50HINO OHGANt/ A f lVN - NAME AND ADOHL Sb Of PaHC, type " Same as ato#, it contractor, provKte NHC Division, Office o* Heg.on US Ntanear Regulatory Commesse and maihng nat-ess }

Same as 8, above.

10 SUPPt.LMLNI ARY NO T E6

<e 11 ADSTRACI (200 w<wds or less)

This journal includes all formal reports m the NUREG series prepared by the NRC staff and contractors; pmceed-ings of conferences and workshops; as well as international agreement reports. The entries in this compilation are indexed for access by title and abstract, secondary report number, personal author, subject. NRC organization for staff and international agreements, contractor, international organuation, and licensed facility.

13 AVAtt. Antul Y STAT EVENT

12. KEY WCHDS/Or SCH:PIOHS (Lat uds u* ptrases th.it wtH awst maearctiers in kxatmg the report )

Unlimited

14. $Er gH6TY CL AS$!FICATICN compilation abstract index Unclassified (ihn keport)

Unclass:fied 1$ NUMtiLH Uf P AGE S t ti MH sC t..

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3 Contractor Index 3

b i:

m*

aC>h" Intnrnational Organization g

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Index

55-9

,e Licensed Facility Index

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